Science.gov

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. 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.

  4. 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)

  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. 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

  8. 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.

  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. 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)

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  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. Fracture Resistance of a Zirconium Alloy with Reoriented Hydrides

    NASA Astrophysics Data System (ADS)

    Chan, Kwai S.; He, Xihua; Pan, Yi-Ming

    2015-01-01

    Zirconium alloy cladding materials typically contain circumferential hydrides that may be reoriented to align along the radial direction when the cladding tubes are heated above and then cooled below the solvus temperature. The objectives of this study were to investigate the critical stress levels required to cause hydride reorientation (HRT) and to characterize the fracture resistance of Zircaloy-2 after hydride reorientation. HRT heat-treatment was performed on hydrogen-charged Zircaloy-2 specimens at 593 K (320 °C) or 623 K (350 °C) for 1 to 2 hours, followed by cooling to 473 K (200 °C). Fracture testing was conducted on hydride-reoriented three-point bend specimens at 473 K (200 °C) using an in situ loading stage inside a scanning electron microscope. Direct observations indicated that the reoriented hydrides, which ranged from ≈1 to 22 μm in lengths, were more prone to fracture at larger sizes (>10 μm) compared to smaller sizes (<0.5 μm). The reoriented hydrides reduced fracture resistance through a void nucleation, growth, and coalescence process at the crack tip. The resulting crack-resistance curves for Zircaloy-2 with reoriented hydrides decrease from 38 to 21 MPa(m)1/2 with increasing hydrogen contents from 51 to 1265 wt ppm hydrogen.

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Characterization of hydrides and delayed hydride cracking in zirconium alloys

    NASA Astrophysics Data System (ADS)

    Fang, Qiang

    This thesis tries to fill some of the missing gaps in the study of zirconium hydrides with state-of-art experiments, cutting edge tomographical technique, and a novel numerical algorithm. A new hydriding procedure is proposed. The new anode material and solution combination overcomes many drawbacks of the AECLRTM hydriding method and leads to superior hydriding result compared to the AECL RTM hydriding procedure. The DHC crack growth velocity of as-received Excel alloy and Zr-2.5Nb alloy together with several different heat treated Excel alloy samples are measured. While it already known that the DHC crack growth velocity increases with the increase of base metal strength, the finding that the transverse plane is the weaker plane for fatigue crack growth despite having higher resistance to DHC crack growth was unexpected. The morphologies of hydrides in a coarse grained Zircally-2 sample have been studied using synchrotron x-rays at ESRF with a new technique called Diffraction Contrast Tomography that uses simultaneous collection of tomographic data and diffraction data to determine the crystallographic orientation of crystallites (grains) in 3D. It has been previously limited to light metals such as Al or Mg (due to the use of low energy x-rays). Here we show the first DCT measurements using high energy x-rays (60 keV), allowing measurements in zirconium. A new algorithm of a computationally effcient way to characterize distributions of hydrides - in particular their orientation and/or connectivity - has been proposed. It is a modification of the standard Hough transform, which is an extension of the Hough transform widely used in the line detection of EBSD patterns. Finally, a basic model of hydrogen migration is built using ABAQUS RTM, which is a mature finite element package with tested modeling modules of a variety of physical laws. The coupling of hydrogen diffusion, lattice expansion, matrix deformation and phase transformation is investigated under

  4. 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.

  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. 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.

  7. 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.

  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. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Design studies of the moderated thermionic heat pipe reactor (MOHTR) concept

    NASA Astrophysics Data System (ADS)

    Ranken, William A.; Turner, John A.

    Design studies, based primarily on neutronics analysis, have been conducted on a thermionic reactor concept that uses a combined beryllium and zirconium hydride moderator to facilitate the incorporation of heat pipe cooling into compact thermionic fuel element (TFE) based designs useful in the tens of kilowatts electrical power regime. The goal of the design approach is to achieve a single point failure free system with technologies such as TFEs, high-temperature heat pipes, and ZrH moderation, which have extensive test databases and have been shown to be capable of long lifetimes. Beryllium is used to thermally couple redundant heat pipes to TFEs and ZrH is added to reduce critical size. Neutronic analysis shows that greater reactivity can be achieved for a given geometry with a combination of the two moderator materials than with ZrH alone and that the combined moderator is much less sensitive to hydrogen loss than more traditional ZrH-moderated thermionic reactor designs. These and other analytical approaches have demonstrated the credibility of a heat pipe cooled thermionic reactor concept that has a reactor height and diameter of 60 cm and a reactor mass of 400 kg for 30-kWe power output.

  14. Design studies of the Moderated Thermonic Heat Pipe Reactor (MOHTR) concept

    SciTech Connect

    Ranken, W.A.; Turner, J.A.

    1991-01-01

    Design studies, based primarily on neutronics analysis, have been conducted on a thermionic reactor concept that uses a combined beryllium and zirconium hydride moderator to facilitate the incorporation of heat pipe cooling into compact thermionic fuel element (TFE) based designs useful in the tens of kilowatts electrical power regime. The goal of the design approach is to achieve a single point failure free system with technologies such as TFEs, high-temperature heat pipes, and ZrH moderation, which have extensive test data bases and have been shown to be capable of long lifetimes. Beryllium is used to thermally couple redundant heat pipes to TFEs and ZrH is added to reduce critical size. Neutronic analysis undertaken to investigate this design approach shows that greater reactivity can be achieved for a given geometry with a combination of the two moderator materials than with ZrH alone and that the combined moderator is much less sensitive to hydrogen loss than more traditional ZrH-moderated thermionic reactor designs. These and other analytical approaches have demonstrated the credibility of a heat pipe cooled thermionic reactor concept that has a reactor height and diameter of 60 cm and a reactor mass of 400 kg for 30-kWe power output. 14 refs., 8 figs.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. 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)

  2. Determination of fracture strength of δ-zirconium hydrides embedded in zirconium matrix at high temperatures

    NASA Astrophysics Data System (ADS)

    Kubo, T.; Kobayashi, Y.; Uchikoshi, H.

    2013-04-01

    The fracture strength of δ-zirconium hydrides embedded in a zirconium matrix was determined at temperatures between 25 °C and 250 °C by ring tensile tests using Zircaloy-2 tubes. Essentially all of the present hydrides in the tubes were re-oriented in the radial direction by a temperature cycling treatment and then tensile stress was applied perpendicular to the hydrides to ensure that brittle fracture would occur at the hydrides. The hydrides failed in a brittle manner below 100 °C where-as the zirconium matrix itself underwent ductile fracture without hydride cracking at temperatures above 200 °C under plane stress condition. Brittle fracture of the hydrides continued to occur at temperatures up to 250 °C under plane strain condition, suggesting that the upper limit temperature for hydride fracture, Tupper, was raised by the triaxial stress state under the plane strain condition. The apparent fracture strength of the hydrides, σhydridef, was determined at temperatures below Tupper from the measured fracture strength of the tubes, making a correction for the compressive transformation stress in the hydrides. σhydridef was about 710 MPa at temperatures between 25 °C and 250 °C at both plane stress and plane strain conditions. The temperature dependency was very small in this temperature range. Tupper was almost equivalent to the cross-over temperature between σhydridef and the ultimate tensile strength (UTS), which suggests that, at temperatures above Tupper, the zirconium matrix would undergo ductile fracture before the stress in the hydride is raised above σhydridef, since UTS is smaller than σhydridef.

  3. 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

  4. 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.

  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. Comparison of delayed hydride cracking behavior of two zirconium alloys

    NASA Astrophysics Data System (ADS)

    Ponzoni, L. M. E.; Mieza, J. I.; De Las Heras, E.; Domizzi, G.

    2013-08-01

    Delayed hydride cracking (DHC) is an important failure mechanism that may occur in Zr alloys during service in water-cooled reactors. Two conditions must be attained to initiate DHC from a crack: the stress intensity factor must be higher than a threshold value called KIH and, hydrogen concentration must exceed a critical value. Currently the pressure tubes for CANDU reactor are fabricated from Zr-2.5Nb. In this paper the critical hydrogen concentration for DHC and the crack velocity of a developmental pressure tube, Excel, was evaluated and compared with that of Zr-2.5Nb. The DHC velocity values measured in Excel were higher than usually reported in Zr-2.5Nb. Due to the higher hydrogen solubility limits in Excel, its critical hydrogen concentration for DHC initiation is 10-50 wppm over that of Zr-2.5Nb in the range of 150-300 °C.

  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. 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.

  10. 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.

  11. 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

  12. Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın

    2011-07-01

    Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

  13. Moderated heat pipe thermionic reactor (MOHTR) module development and test

    NASA Astrophysics Data System (ADS)

    Merrigan, Michael A.; Trujillo, Vincent L.

    1992-01-01

    The Moderated Heat Pipe Thermionic Reactor (MOHTR) thermionic space reactor design combines the low risk technology associated with the Thermionic Fuel Element (TFE) Verification Program with the high reliability heat transfer capability of liquid metal heat pipe technology. The resulting design concept, capable of implementation over the power range of 10 to 100 kWe, offers efficiency and reliability with reduced risk of single point failures. The union of TFE and heat pipe technology is achieved by imbedding TFEs and heat pipes in a beryllium matrix to which they are thermally coupled by brazing or by liquid metal (NaK or Na) bonding. The reactor employs an array of TFE modules, each comprising a TFE, a zirconium hydride (ZrH) cylinder for neutron moderation, and heat pipes for transport of heat from the collector surface of the TFE to the waste heat radiator. An advantage of the design is the low temperature drop from the collector surface to the radiating surface. This is a result of the elimination of electrical insulation from the heat transport path through electrical isolation of the modules. The module used in this study consisted of a beryllium core, and electrical cartridge heater simulating the TFE, and three heat pipes to dissipate the waste heat. The investigation was focused on the thermal performance of the assembly, including evaluation of the sodium and braze bonding options for minimizing the thermal resistance between the elements, the temperature distribution in the beryllium matrix, and the heat pipe performance. Continuing subjects of the investigation include performance of the heat pipes through start-up transients, during normal operation, and in a single heat pipe failure mode. Secondary objectives of the investigation include correlation of analytic models for the thermionic element and module including the effects of gap thermal conductances at the modules electrically insulated surfaces.

  14. 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.

  15. 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

  16. 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.

  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. 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.

  19. 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

  20. 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.

  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. 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)

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

    DOE PAGES

    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

  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. 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

  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. Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

    NASA Astrophysics Data System (ADS)

    Rico, A.; Martin-Rengel, M. A.; Ruiz-Hervias, J.; Rodriguez, J.; Gomez-Sanchez, F. J.

    2014-09-01

    It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young's modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200 ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young's modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200 ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found.

  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. 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)

  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. COOLED NEUTRONIC REACTOR

    DOEpatents

    Binner, C.R.; Wilkie, C.B.

    1958-03-18

    This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

  15. 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)

  16. 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.

  17. 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.

  18. Space power reactor in-core thermionic multicell evolutionary (S-prime) design

    NASA Astrophysics Data System (ADS)

    Determan, William R.; Van Hagan, Tom H.

    1993-01-01

    A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

  19. Space power reactor in-core thermionic multicell evolutionary (S-prime) design

    SciTech Connect

    Determan, W.R. ); Van Hagan, T.H. )

    1993-01-20

    A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m[sup 2] heat pipe space radiator.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. NMR study of hydrogen diffusion in zirconium hydride

    SciTech Connect

    Korn, C.; Goren, S.D.

    1986-01-01

    The nuclear-magnetic-resonance method was used to study the diffusion of hydrogen in zirconium hydride by measuring the temperature dependence of T/sub 1/ in a temperature range where the major relaxation mechanism was due to hydrogen diffusion. The samples investigated were ZrH/sub 1.588/, ZrH/sub 1.629/, ZrH/sub 1.684/, ZrH/sub 1.736/, ZrH/sub 1.815/, ZrH/sub 1.910/, and ZrH/sub 1.960/. These spanned both the cubic and tetragonal phases. The activation energy was found to be independent of hydrogen concentration in the cubic phase with E/sub a/ = 13.4 +- 0.4 kcal/mol and a preexponential factor given by A = (1/2)(2-x)(45 +- 10) x 10/sup 12/ Hz. In the tetragonal phase the activation energy of the bulk of the hydrogen increased modestly with concentration. In addition, it was discovered that a new very fast hydrogen channel was created by the tetragonality for approx.3% of the hydrogen. They jump with a preexponential factor that is about 2 orders of magnitude larger than that of the rest of the hydrogen. A comparison was also made between the Bloembergen-Purcell-Pound, the Barton-Sholl, and the Bustard theories for nuclear magnetic relaxation due to diffusion.

  5. NMR study of hydrogen diffusion in zirconium hydride

    NASA Astrophysics Data System (ADS)

    Korn, C.; Goren, S. D.

    1986-01-01

    The nuclear-magnetic-resonance method was used to study the diffusion of hydrogen in zirconium hydride by measuring the temperature dependence of T1 in a temperature range where the major relaxation mechanism was due to hydrogen diffusion. The samples investigated were ZrH1.588, ZrH1.629, ZrH1.684, ZrH1.736, ZrH1.815, ZrH1.910, and ZrH1.960. These spanned both the cubic and tetragonal phases. The activation energy was found to be independent of hydrogen concentration in the cubic phase with Ea=13.4+/-0.4 kcal/mol and a preexponential factor given by A=(1/2)(2-x)(45+/-10)×1012 Hz. In the tetragonal phase the activation energy of the bulk of the hydrogen increased modestly with concentration. In addition, it was discovered that a new very fast hydrogen channel was created by the tetragonality for ~3% of the hydrogen. They jump with a preexponential factor that is about 2 orders of magnitude larger than that of the rest of the hydrogen. A comparison was also made between the Bloembergen-Purcell-Pound, the Barton-Sholl, and the Bustard theories for nuclear magnetic relaxation due to diffusion.

  6. 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)

  7. 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.

  8. 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.

  9. 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)

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

    DOE PAGES

    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

  11. 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.

  12. 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.

  13. 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.

  14. 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)

  15. The computational-and-experimental investigation into the head-flow characteristic of the two-stage ejector for the emergency core cooling system of the NPP with a water-moderated water-cooled power reactor

    NASA Astrophysics Data System (ADS)

    Parfenov, Yu. V.

    2013-09-01

    The results of the computational-and-experimental investigation into the two-stage ejector for the emergency cooling system of the core of the water-moderated water-cooled power reactor. The results of experimental investigations performed for the ejector model at the JSC "EREC" and the result of calculations performed using the REMIX CFD code are presented.

  16. 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.

  17. 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.

  18. 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)

  19. 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.

  20. Zirconium alloy performance in light water reactors: A review of UK and Scandinavian experience

    SciTech Connect

    Pickman, D.O.

    1994-12-31

    Various aspects of zirconium alloy development for light water reactors in the UK and Scandinavia are reviewed, including the contribution made by some unique nuclear testing facilities. Among the problems encountered were the irradiation enhancement of corrosion and hydrogen pickup, crud deposition, iodine-induced stress-corrosion cracking on power ramping, and severe cladding deformation in loss-of-coolant accident conditions. The causes and behavior of defects, including hydride defects and fretting corrosion, are discussed.

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. Multiphysics phase field modeling of hydrogen diffusion and delta-hydride precipitation in alpha-zirconium

    NASA Astrophysics Data System (ADS)

    Jokisaari, Andrea M.

    Hydride precipitation in zirconium is a significant factor limiting the lifetime of nuclear fuel cladding, because hydride microstructures play a key role in the degradation of fuel cladding. However, the behavior of hydrogen in zirconium has typically been modeled using mean field approaches, which do not consider microstructural evolution. This thesis describes a quantitative microstructural evolution model for the alpha-zirconium/delta-hydride system and the associated numerical methods and algorithms that were developed. The multiphysics, phase field-based model incorporates CALPHAD free energy descriptions, linear elastic solid mechanics, and classical nucleation theory. A flexible simulation software implementing the model, Hyrax, is built on the Multiphysics Object Oriented Simulation Environment (MOOSE) finite element framework. Hyrax is open-source and freely available; moreover, the numerical methods and algorithms that have been developed are generalizable to other systems. The algorithms are described in detail, and verification studies for each are discussed. In addition, analyses of the sensitivity of the simulation results to the choice of numerical parameters are presented. For example, threshold values for the CALPHAD free energy algorithm and the use of mesh and time adaptivity when employing the nucleation algorithm are studied. Furthermore, preliminary insights into the nucleation behavior of delta-hydrides are described. These include a) the sensitivities of the nucleation rate to temperature, interfacial energy, composition and elastic energy, b) the spatial variation of the nucleation rate around a single precipitate, and c) the effect of interfacial energy and nucleation rate on the precipitate microstructure. Finally, several avenues for future work are discussed. Topics encompass the terminal solid solubility hysteresis of hydrogen in zirconium and the effects of the alpha/delta interfacial energy, as well as thermodiffusion, plasticity

  6. 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.

  7. 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.

  8. 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.

  9. 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)

  10. 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)

  11. 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.

  12. 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.

  13. Zirconium hydrides and Fe redistribution in Zr-2.5%Nb alloy under ion irradiation

    NASA Astrophysics Data System (ADS)

    Idrees, Y.; Yao, Z.; Cui, J.; Shek, G. K.; Daymond, M. R.

    2016-11-01

    Zr-2.5%Nb alloy is used to fabricate the pressure tubes of the CANDU reactor. The pressure tube is the primary pressure boundary for coolant in the CANDU design and is susceptible to delayed hydride cracking, reduction in fracture toughness upon hydride precipitation and potentially hydride blister formation. The morphology and nature of hydrides in Zr-2.5%Nb with 100 wppm hydrogen has been investigated using transmission electron microscopy. The effect of hydrides on heavy ion irradiation induced decomposition of the β phase has been reported. STEM-EDX mapping was employed to investigate the distribution of alloying elements. The results show that hydrides are present in the form of stacks of different sizes, with length scales from nano- to micro-meters. Heavy ion irradiation experiments at 250 °C on as-received and hydrided Zr-2.5%Nb alloy, show interesting effects of hydrogen on the irradiation induced redistribution of Fe. It was found that Fe is widely redistributed from the β phase into the α phase in the as-received material, however, the loss of Fe from the β phase and subsequent precipitation is retarded in the hydrided material. This preliminary work will further the current understanding of microstructural evolution of Zr based alloys in the presence of hydrogen.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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

  20. 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

  1. Crystal chemistry of sodium zirconium phosphate based simulated ceramic waste forms of effluent cations (Ba(2+), Sn(4+), Fe(3+), Cr(3+), Ni(2+) and Si(4+)) from light water reactor fuel reprocessing plants.

    PubMed

    Shrivastava, O P; Chourasia, Rashmi

    2008-05-01

    A novel concept of immobilization of light water reactor (LWR) fuel reprocessing waste effluent through interaction with sodium zirconium phosphate (NZP) has been established. Such conversion utilizes waste materials like zirconium and nickel alloys, stainless steel, spent solvent tri-butyl phosphate and concentrated solution of NaNO(3). The resultant multi component NZP material is a physically and chemically stable single phase crystalline product having good mechanical strength. The NZP matrix can also incorporate all types of fission product cations in a stable crystalline lattice structure; therefore, the resultant solid solutions deserve quantification of crystallographic data. In this communication, crystal chemistry of the two types of simulated waste forms (type I-Na(1.49)Zr(1.56)Sn(0.02)Fe(0).(28)Cr(0.07)Ni(0.07)P(3)O(12) and type II-Na(1.35)Ba(0.14)Zr(1.56)Sn(0.02)Fe(0).(28)Cr(0.07)Ni(0.07)P(2.86)Si(0.14)O(12)) has been investigated using General Structure Analysis System (GSAS) programming of the X-ray powder diffraction data. About 4001 data points of each have been subjected to Rietveld analysis to arrive at a satisfactory structural convergence of Rietveld parameters; R-pattern (R(p))=0.0821, R-weighted pattern (R(wp))=0.1266 for type I and R(p)=0.0686, R(wp)=0.0910 for type II. The structure of type I and type II waste forms consist of ZrO(6) octahedra and PO(4) tetrahedra linked by the corners to form a three-dimensional network. Each phosphate group is on a two-fold rotation axis and is linked to four ZrO(6) octahedra while zirconium octahedra lies on a three-fold rotation axis and is connected to six PO(4) tetrahedra. Though the expansion along c-axis and shrinkage along a-axis with slight distortion of bond angles in the synthesized crystal indicate the flexibility of the structure, the waste forms are basically of NZP structure. Morphological examination by SEM reveals that the size of almost rectangular parallelepiped crystallites varies

  2. 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

  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. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Advanced sodium fast reactor accident source terms :

    SciTech Connect

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; 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 event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions 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.

  10. 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

  11. 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.

  12. A Probabilistic Integrity Assessment of Flaw in Zirconium Alloy Pressure Tube Considering Delayed Hydride Cracking

    NASA Astrophysics Data System (ADS)

    Kwak, Sang Log; Lee, Joon Seong; Kim, Young Jin; Park, Youn Won

    In the CANDU nuclear reactor, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the nuclear fuel bundles and heavy water coolant. Pressure tubes are major component of nuclear reactor, but only selected samples are periodically examined due to numerous numbers of tubes. Pressure tube material gradually pick up deuterium, as such are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC), which is the characteristic of pressure tube integrity evaluation. If cracks are not detected, such a cracking mechanism could lead to unstable rupture of the pressure tube. Up to this time, integrity evaluations are performed using conventional deterministic approaches. So it is expected that the results obtained are too conservative to perform a rational evaluation of lifetime. In this respect, a probabilistic safety assessment method is more appropriate for the assessment of overall pressure tube safety. This paper describes failure criteria for probabilistic analysis and fracture mechanics analyses of the pressure tubes in consideration of DHC. Major input parameters such as initial hydrogen concentration, the depth and aspect ratio of an initial surface crack, DHC velocity and fracture toughness are considered as probabilistic variables. Failure assessment diagram of pressure tube material is proposed and applied in the probabilistic analysis. In all the analyses, failure probabilities are calculated using the Monte Carlo simulation. As a result of analysis, conservatism of deterministic failure criteria is showed.

  13. Sodium fast reactor evaluation: Core materials

    NASA Astrophysics Data System (ADS)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

  14. 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%.

  15. A phase-field model to study the effects of temperature change on shape evolution of γ-hydrides in zirconium

    NASA Astrophysics Data System (ADS)

    Bair, Jacob; Asle Zaeem, Mohsen; Tonks, Michael

    2016-10-01

    A temperature-dependent phase-field model is developed to study the effects of temperature change on shape evolution of γ-hydrides in an α-zirconium matrix. To construct the temperature-dependent free energy functional of the phase-field model, Gibbs free energies of formation from previous experiments are employed, and one conserved and three non-conserved phase-field variables are used for hydrogen concentration and hydride orientations, respectively. The mixed order evolution equations of phase-field variables coupled with mechanical equilibrium equations are solved in a finite element framework. Results from isothermal simulations of seeded and random nucleation in single crystal α-zirconium matrix show that the thickness of non-equilibrium hydrides varies with temperature during evolution, and the hydrides are more rod-like (thinner) at higher temperatures and thicker at lower temperatures. Quench simulations with random nucleation indicate that the majority of precipitation occurs at early stages of quenching, but the size and shape of hydrides change as the temperature decreases. Simulations from random nucleation of hydrides in a polycrystalline α-zirconium matrix show a higher concentration of precipitates along high angle grain boundaries.

  16. 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)

  17. 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.

  18. 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.

  19. 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.

  20. Polarization of modified titanium and titanium-zirconium creates nano-structures while hydride formation is modulated

    NASA Astrophysics Data System (ADS)

    Frank, Matthias J.; Walter, Martin S.; Bucko, Miroslaw M.; Pamula, Elzbieta; Lyngstadaas, S. Petter; Haugen, Håvard J.

    2013-10-01

    The majority of titanium based bone-level dental implants available on the market today feature a sand-blasted and acid-etched (SBAE) surface that contains comparably high hydrogen levels. Cathodic polarization of titanium in acidic solutions is known to further increase titanium hydride on the surface. The aim of this study was to explore the effect of cathodic reduction on titanium (Ti) and titanium-zirconium (TiZr) with a SBAE surface in order to investigate the potential of such a process for further improving surfaces for bone anchored dental implants. Samples of both materials were cathodically polarized in acidic solution at different current densities and for different process times. Chemical analysis of the hydrogen levels by SIMS showed that cathodic reduction re-arranged the hydride already present on the surfaces from the etching process but could not significantly increase hydride levels. The hydrogen layer created by the preceding hot acid etching appeared to modulate further hydride creation. Analysis of the surface topography by SEM showed changes to the nano-topography of both materials after polarization. TiZr showed homogeneously distributed nano-spheres as they were already observed for TiZr SBAE at increased size of 80-100 nm on the whole surface. By contrast, polarization of Ti created nano-nodules and nano-spheres of 150-200 nm on the surface. These spheres were interconnected to form flower-like structures along the ridges and peaks of the surface. Moreover the flanks were covered by a rippled structure of isotropically distributed small-diameter (10-20 nm) nano-nodules.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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

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

    PubMed

    Nayak, Nadiya B; Nayak, Bibhuti B

    2016-03-16

    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.

  8. 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.

  9. 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.

  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. 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.

  12. 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.

  13. 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.

  14. Metal hydrides reactors with improved dynamic characteristics for a fast cycling hydrogen compressor

    NASA Astrophysics Data System (ADS)

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

    2009-08-01

    This paper presents an investigation of coupled heat and mass transfer process in metal hydrides hydrogen storage reactors. Hydrogen storage and compression performance of our designed and developed reactors are studied by varying the operating parameters and analyzing the effects of metal hydride bed parameters. The metal alloy selected to characterize the cycling behaviour of reactors is LaNi5, material synthesized and characterized by us in the range 20-80°C. Four types of metal hydride reactors were tested with the aim to provide a fast hydrogen absorption-desorption cycle, able to be thermally cycled at rapid rates. Some new technical solutions have been studied to make a step forward in reducing the duration of the reactors cycle, which combines the effective increase of the thermal conductivity and good permeability to hydrogen gas. Dynamic characteristic of developed fast metal hydride reactors is improved using our novel mixture metal hydride-CA conductive additive due to the increased effective thermal conductivity of the alloy bed. The advanced hydride bed design with high heat transfer capabilities can be thermally cycled at a rapid rate, under 120 seconds, in order to process high hydrogen flow rates.

  15. 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.

  16. ExoMol molecular line lists - X. The spectrum of sodium hydride

    NASA Astrophysics Data System (ADS)

    Rivlin, Tom; Lodi, Lorenzo; Yurchenko, Sergei N.; Tennyson, Jonathan; Le Roy, Robert J.

    2015-07-01

    Accurate and complete rotational, rotational-vibrational and rotational-vibrational-electronic line lists are calculated for sodium hydride: both the NaH and NaD isotopologues are considered. These line lists cover all ro-vibrational states of the ground (X 1Σ+) and first excited (A 1Σ+) electronic states. The calculations use available spectroscopically-determined potential energy curves and new high-quality, ab initio dipole moment curves. Partition functions for both isotopologues are calculated and the effect of quasi-bound states is considered. The resulting line lists are suitable for temperatures up to about 7000 K and are designed for studies of exoplanet atmospheres, brown dwarfs and cool stars. In particular, the NaH A - X band is found to show a broad absorption feature at about 385 nm which should provide a signature for the molecule. All partition functions, lines and transitions are available as supplementary information to this article and at www.exomol.com.

  17. 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.

  18. Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator

    NASA Astrophysics Data System (ADS)

    Sarchami, Araz

    A three-dimensional numerical modeling of the thermo hydraulics of Canadian Deuterium Uranium (CANDU) nuclear reactor is conducted. The moderator tank is a Pressurized heavy water reactor which uses heavy water as moderator in a cylindrical tank. The main use of the tank is to bring the fast neutrons to the thermal neutron energy levels. The moderator tank compromises of several bundled tubes containing nuclear rods immersed inside the heavy water. It is important to keep the water temperature in the moderator at sub-cooled conditions, to prevent potential failure due to overheating of the tubes. Because of difficulties in measuring flow characteristics and temperature conditions inside a real reactor moderator, tests are conducted using a scaled moderator in moderator test facility (MTF) by Chalk River Laboratories of Atomic Energy of Canada Limited (CRL, AECL). MTF tests are conducted using heating elements to heat tube surfaces. This is different than the real reactor where nuclear radiation is the source of heating which results in a volumetric heating of the heavy water. The data recorded inside the MTF tank have shown levels of fluctuations in the moderator temperatures and requires in depth investigation of causes and effects. The purpose of the current investigation is to determine the causes for, and the nature of the moderator temperature fluctuations using three-dimensional simulation of MTF with both (surface heating and volumetric heating) modes. In addition, three dimensional simulation of full scale actual moderator tank with volumetric heating is conducted to investigate the effects of scaling on the temperature distribution. The numerical simulations are performed on a 24-processor cluster using parallel version of the FLUENT 12. During the transient simulation, 55 points of interest inside the tank are monitored for their temperature and velocity fluctuations with time.

  19. 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.

  20. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    NASA Astrophysics Data System (ADS)

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.

    2013-07-01

    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

  1. Mass estimates of very small reactor cores fueled by Uranium-235, U-233 and Cm-245

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Lipinski, Ronald J.

    2001-02-01

    This paper explores the possibility of manufacturing very small reactors from U-235, U-233 and Cm-245. Pin type reactor systems fueled with uranium or curium metal zirconium hydride (UZrH or CmZrH) are compared with similar designs using U-235. Criticality measurements of homogeneous water uranium systems, suggest that reactor subsystem masses have a broad minimum for hydrogen-to-uranium atom ratios that vary from 25-250. This paper compares the masses of metal-hydride fueled reactor systems that use U-235, U-233, and Cm-245 fuel with hydrogen-to-metal atom ratios from 20-300 when cooled by gas (HeXe), liquid metal (Na), and water. The results indicate that water cooled reactors in general have the smallest reactor subsystem mass. For gas and liquid-metal cooled reactors U-233 subsystems have total masses that are about 1/2 those of similarly designed U-235 fuel reactors. Reactor subsystems consisting of 11.2% enriched Cm-245 (balance Cm-244) that can be obtained from fuel reprocessing have system masses comparable to that of U-233. The smallest reactor subsystem masses were on the order of 60-80 kg for U-233 fueled water cooled reactors. .

  2. 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.

  3. 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.

  4. 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)

  5. 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.

  6. 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.

  7. Space-reactor system and subsystem investigations: cost and schedule estimates for reactor and shield subsystems technology development. SP-100 Program

    SciTech Connect

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

    1983-06-30

    This report presents cost and schedule estimates of the technology development for reactor and shielding subsystems of a 100-kWe class space reactor electric system. 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 reactor; 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 in this report would be used in conjunction with unmanned payloads.

  8. 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)

  9. The influence of moderate reduction in dietary sodium on human salivary sodium concentration.

    PubMed

    Christensen, C M; Bertino, M; Beauchamp, G K; Navazesh, M; Engelman, K

    1986-01-01

    Twenty-four healthy subjects were placed for 12-13 weeks on diets that reduced average sodium intake from 145 to 74 m-equiv. Na+/day as determined by multiple 24-h urine collections before and during the diet. Whole-mouth resting and stimulated saliva was collected and analysed for flow rate and sodium concentration several times before and during the low-sodium period. Sodium restriction did not influence salivary flow rates but salivary sodium levels fell 25 per cent for resting and 17 per cent for stimulated saliva. Thus moderate reductions in sodium intake are accompanied by significantly lower salivary sodium levels.

  10. 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.

  11. Immobilisation of molybdenum from fuel reprocessing wastes into sodium zirconium phosphate ceramics

    NASA Astrophysics Data System (ADS)

    Pet'kov, V. I.; Sukhanov, M. V.

    2003-01-01

    A new possibility for the incorporation of molybdenum in sodium zirconium phosphate (NZP) nuclear waste form was examined. The existence of molybdenum containing NZP-type structures was assessed on the basis of crystal chemical principles. New data were deduced for the system Na1-xZr2(PO4)3-x(MoO4)x (0≤x≤1). The stability region for this solid solution having a NZP structure was determined. It was found that molybdenum can be incorporation into NZP-type ceramics without significant changes of the three-dimensional framework structure.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Estimation of the sub-criticality of the sodium-cooled fast reactor Monju using the modified neutron source multiplication method

    SciTech Connect

    Truchet, G.; Van Rooijen, W. F. G.; Shimazu, Y.; Yamaguchi, K.

    2012-07-01

    The Modified Neutron Source Method (MNSM) is applied to the Monju reactor. This static method to estimate sub-criticality has already given good results on commercial Pressurized Water Reactors. The MNSM consists both in the extraction of the fundamental mode seen by a detector to avoid the effect of higher modes near sources, and the correction of flux distortion effects due to control rod movement. Among Monju's particularities that have a big influence on MNSM factors are: the presence of two californium sources and the position of the detector which is located far from the core outside of the reactor vessel. The importance of spontaneous fission and ({alpha}, n) reactions which have increased during the shutdown period of 15 years will also be discussed. The relative position of detectors and sources deeply affect the correction factors in some regions. In order to evaluate the detector count rate, an analytical propagation has been conducted from the reactor vessel. For two subcritical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). (authors)

  18. 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.

  19. System startup simulation for an in-core thermionic reactor with heat pipe cooling

    NASA Astrophysics Data System (ADS)

    Determan, William R.; Otting, William D.

    1992-01-01

    The heat pipe cooled thermionic (HPTI) reactor relies on in-core sodium heat pipes to provide a redundant means of cooling the 72 thermionic fuel elements (TFEs) which comprise the 40-kWe reactor core assembly. In-core heat pipe cooling was selected for the reactor design due to a requirement for multiple system on-orbit restarts over its lifetime. Powering up the reactor requires the in-core and radiator heat pipes to undergo a thaw cycle with a rapid ascension in power to their operating temperatures. The present study considers how fast the thaw-out and power ascension cycle can be safely accomplished within a reactor core. As part of the study, a transient startup simulator model of the heat pipe cooled reactor system was developed. Results of the startup transient simulation are provided.

  20. 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.

  1. 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

  2. Liquid-metal-cooled reactor

    DOEpatents

    Hutter, Ernest

    1982-01-01

    A perforated depressor plate extending across the bottom of the instrument ree 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.

  3. 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.

  4. 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.

  5. Heat pipe cooled thermionic reactor core fabrication

    NASA Astrophysics Data System (ADS)

    Harlan Horner, M.; Van Hagan, Thomas H.; Determan, William R.

    1992-01-01

    Thermionic and driver fuel elements in an in-core heat pipe cooled reactor will reject heat to a surrounding array of redundant heat pipes. Such structures present a formidable fabrication problem if approached conventionally through assembly of tubular structures. A reasonable method for fabrication of such reactor cores is described in this study. The technique selected involves the use of hot isostatic processing to minimize the amount of material in heat pipe walls, maximize available vapor flow cross sectional area and provide accurate location of fuel elements.

  6. 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.

  7. 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)

  8. 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.

  9. CALIOP: a multichannel design code for gas-cooled fast reactors. Code description and user's guide

    SciTech Connect

    Thompson, W.I.

    1980-10-01

    CALIOP is a design code for fluid-cooled reactors composed of parallel fuel tubes in hexagonal or cylindrical ducts. It may be used with gaseous or liquid coolants. It has been used chiefly for design of a helium-cooled fast breeder reactor and has built-in cross section information to permit calculations of fuel loading, breeding ratio, and doubling time. Optional cross-section input allows the code to be used with moderated cores and with other fuels.

  10. 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

  11. 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

  12. 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).

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  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. 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.

  19. 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.

  20. 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

  1. 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.

  2. Liquid metal cooled reactors for space power applications

    NASA Technical Reports Server (NTRS)

    Bailey, S.; Vaidyanathan, S.; Van Hoomissen, J.

    1985-01-01

    The technology basis for evaluation of liquid metal cooled space reactors is summarized. Requirements for space nuclear power which are relevant to selection of the reactor subsystem are then reviewed. The attributes of liquid metal cooled reactors are considered in relation to these requirements in the areas of liquid metal properties, neutron spectrum characteristics, and fuel form. Key features of typical reactor designs are illustrated. It is concluded that liquid metal cooled fast spectrum reactors provide a high confidence, flexible option for meeting requirements for SP-100 and beyond.

  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. 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.

  5. 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.

  6. 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.

  7. 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

  8. A Qualitative Reactivity and Isotopic Assessment of Fuels for Lead-Alloy Cooled Fast Reactors

    SciTech Connect

    Kevan Weaver; Phil MacDonald

    2004-09-01

    Various methods have been proposed to transmute and thus consume the current inventory of transuranic waste from spent light water reactor (LWR) fuel and plutonium from weapons. We discuss the neutronics performance of nonfertile, fertile metallic, and fertile nitride fuels loaded with 20 to 30 wt% LWR-grade plutonium plus minor actinides and burned in an open-lattice lead-alloy-cooled fast reactor, with an emphasis on the fuel cycle life and spent fuel isotopic content. As a comparison, similar fuel was also studied in a sodium-cooled fast reactor. Our calculations show that the average actinide burn rate for fertile-free fuel 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. In addition, our calculations show that the effective full-power days (EFPDs) of operation (or equivalent reactivity-limited burnup) using fertile fuel can extend beyond 20 yr, and the average actinide burn rate is similar for both the sodium- and lead-bismuth-cooled cases, ranging from 0.5 to 0.9 g/MWd. Using the same parameters (i.e., a large pitch-to-diameter ratio, same linear power, and fissile/fertile loading, etc.), the lead-alloy-cooled cases had an EFPD that was 18% to several times greater than their sodium-cooled counterparts. However, tight sodium-cooled lattices are equivalent to the looser lead-alloy lattices in terms of beginning-of-life excess reactivity.

  9. A Qualitative Reactivity and Isotopic Assessment of Fuels for Lead-Alloy-Cooled Fast Reactors

    SciTech Connect

    Weaver, Kevan D.; MacDonald, Philip E.

    2004-09-15

    Various methods have been proposed to transmute and thus consume the current inventory of transuranic waste from spent light water reactor (LWR) fuel and plutonium from weapons. We discuss the neutronics performance of nonfertile, fertile metallic, and fertile nitride fuels loaded with 20 to 30 wt% LWR-grade plutonium plus minor actinides and burned in an open-lattice lead-alloy-cooled fast reactor, with an emphasis on the fuel cycle life and spent fuel isotopic content. As a comparison, similar fuel was also studied in a sodium-cooled fast reactor. Our calculations show that the average actinide burn rate for fertile-free fuel 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. In addition, our calculations show that the effective full-power days (EFPDs) of operation (or equivalent reactivity-limited burnup) using fertile fuel can extend beyond 20 yr, and the average actinide burn rate is similar for both the sodium- and lead-bismuth-cooled cases, ranging from 0.5 to 0.9 g/MWd. Using the same parameters (i.e., a large pitch-to-diameter ratio, same linear power, and fissile/fertile loading, etc.), the lead-alloy-cooled cases had an EFPD that was 18% to several times greater than their sodium-cooled counterparts. However, tight sodium-cooled lattices are equivalent to the looser lead-alloy lattices in terms of beginning-of-life excess reactivity.

  10. 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)

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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

  16. 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

  17. Solid methane cold moderator for the IBR-2 reactor

    SciTech Connect

    Beliakov, A. A.; Tretiakov, I. T.; Shabalin, E. P.; Golikov, V. V.; Luschivkov, V I.

    1997-09-01

    The paper describes the research and design work carried out since 1986 at the Frank Laboratory of Neutron Physics of the Joint Institute for Nuclear Research in Dubna to create a cryogenic moderator for the IBR-22 reactor using solid methane as a moderating substance.

  18. 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.

  19. 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)

  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. 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.

  3. 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.

  4. 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.

  5. 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).

  6. 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.

  7. 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

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

    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

  9. 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. .

  10. Coupled Reactor Kinetics and Heat Transfer Model for Heat Pipe Cooled Reactors

    SciTech Connect

    WRIGHT,STEVEN A.; HOUTS,MICHAEL

    2000-11-22

    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). The 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.

  11. Chlorination/dechlorination studies relating to proposed cooling towers for K and C reactors

    SciTech Connect

    Wilde, E.W.

    1986-12-01

    The purpose of this study was to obtain data on the chlorination and dechlorination of the cooling water that is pumped through the C- and K-Reactor heat exchangers. More specifically, data were obtained on: (1) chlorine demand and dissipation rates; (2) the effectiveness of sodium sulfite as a dechlorinating agent; (3) the toxicity of chlorination and dechlorination on indigenous fish; (4) the effects of dechlorination on water quality; and (5) the comparative precision of three conventional residual chlorine detection methods. The study results strongly indicate that implementation of present plans for chlorination and dechlorination (with sodium sulfite) of SRP reactor cooling towers will cause no significant adverse environmental impacts and should not result in problems in meeting future regulatory guidelines for TRC. 9 figs., 11 tabs.

  12. 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.

  13. 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.

  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. 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.

  16. 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.

  17. 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)

  18. 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.

  19. 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.

  20. Moderate-temperature zeolitic alteration in a cooling pyroclastic deposit

    USGS Publications Warehouse

    Levy, S.S.; O'Neil, J.R.

    1989-01-01

    The locally zeolitized Topopah Spring Member of the Paintbrush Tuff (13 Myr.), Yucca Mountain, Nevada, U.S.A., is part of a thick sequence of zeolitized pyroclastic units. Most of the zeolitized units are nonwelded tuffs that were altered during low-temperature diagenesis, but the distribution and textural setting of zeolite (heulandite-clinoptilolite) and smectite in the densely welded Topopah Spring tuff suggest that these hydrous minerals formed while the tuff was still cooling after pyroclastic emplacement and welding. The hydrous minerals are concentrated within a transition zone between devitrified tuff in the central part of the unit and underlying vitrophyre. Movement of liquid and convected heat along fractures from the devitrified tuff to the ritrophyre caused local devitrification and hydrous mineral crystallization. Oxygen isotope geothermometry of cogenetic quartz confirms the nondiagenetic moderate temperature origin of the hydrous minerals at temperatures of ??? 40-100??C, assuming a meteoric water source. The Topopah Spring tuff is under consideration for emplacement of a high-level nuclear waste repository. The natural rock alteration of the cooling pyroclastic deposit may be a good natural analog for repository-induced hydrothermal alteration. As a result of repository thermal loading, temperatures in the Topopah Spring vitrophyre may rise sufficiently to duplicate the inferred temperatures of natural zeolitic alteration. Heated water moving downward from the repository into the vitrophyre may contribute to new zeolitic alteration. ?? 1989.

  1. Utilization/Disposition of Reactor-Grade Plutonium in High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Bonin, Bernard; Greneche, Dominique; Carre, Frank; Damian, Frederic; Doriath, Jean-Yves

    2004-03-15

    High-temperature gas-cooled reactors (HTRs) are able to accommodate a wide variety of mixtures of fissile and fertile materials without any significant modification of the core design. This flexibility is due to an uncoupling between the parameters of cooling geometry and the parameters that characterize neutronic optimization (moderation ratio or heavy nuclide concentration and distribution).Among other advantageous features, an HTR core has a better neutron economy than a light water reactor (LWR) because there is much less parasitic capture in the moderator (capture cross section of graphite is 100 times less than the one of water), in internal structures, and in fission products (because of a harder spectrum).Moreover, thanks to the high strength of the coated particles, HTR fuels are able to reach very high burnups, far beyond the possibilities offered by other fuels (except the special case of molten salt reactors).These features make HTRs potentially interesting for closing the nuclear fuel cycle and stabilizing the plutonium inventory.A large number of fuel cycle studies are already available today on three main categories of fuel cycles involving HTRs: (a) high-enriched uranium cycle, based on thorium utilization as a fertile material and high-enriched uranium as a fissile material; (b) low-enriched uranium cycle (LEU), where only LEU is used (from 5 to 15%); (c) plutonium cycle based on the utilization of plutonium only as a fissile material, with (or without) fertile materials.Plutonium consumption at high burnups in HTRs has already been tested with encouraging results under the DRAGON project and at Peach Bottom. To maximize plutonium consumption, recent core studies have also been performed on plutonium HTR cores, with special emphasis on weapon-grade plutonium consumption. We complete the picture by a core study for an HTR burning reactor-grade plutonium. Limits in burnup due to core neutronics are investigated for this type of core.With these limits

  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. 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.

  4. 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.

  5. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

    In this paper the characteristics of six designs for power levels of 2, 10, and 20 MWe for operating times of 1 and 7 y are described. The operating conditions for these arbitrary designs were chosen to minimize system specific mass. The designs are based on recent work which benefits from earlier analyses of nuclear space power systems conducted at our Laboratory. Both gas- and liquid-cooled reactors had been considered. Pitts and Walter (1970) reported on the results of a detailed study of a 10-MWe lithium-cooled reactor in a potassium Rankine system. Unpublished results (1966) of a computer analysis provide details of an argon-cooled reactor in an argon Brayton system. The gas-cooled reactor design was based on extensive development work on the 500-MWth reactor for the nuclear ramjet (Pluto) as described by Walter (1964). The designs discussed here draw heavily on the Pluto project experience, which culminated in a successful full-power ground test as reported by Reynolds (1964). At higher power levels gas-cooled reactors coupled with Brayton systems with advanced radiator designs become attractive.

  6. 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)

  7. 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

  8. 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.

  9. Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

    SciTech Connect

    Krass, A.W.

    2005-12-19

    This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.

  10. 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.

  11. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1993-07-01

    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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Liquid metal reactor air cooling baffle

    DOEpatents

    Hunsbedt, Anstein

    1994-01-01

    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.

  5. 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;

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

    NASA Astrophysics Data System (ADS)

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

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (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 along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) 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.

  7. Gas-cooled reactor power systems for space

    NASA Astrophysics Data System (ADS)

    Walter, Carl E.

    Large amounts of electric power are required for some of the systems envisioned in support of SDI. Since various applications are being considered, and an overall power architecture study has not been completed, the required power levels and corresponding operating times for specific systems are not known. The characteristics of six designs for power levels of 2, 10 and 20 MWe for operating time of 1 and 7 yrs are described. The operating conditions for these arbitrary designs were chosen to minimize system specific mass. Both gas and liquid cooled reactors are considered. The designs discussed draw heavily on the Pluto project experience. Gas cooled thermal reactors coupled with Brayton cycle power conversion appear to provide reasonable multimegawatt space power systems. An advanced radiation design must be developed which can meet the mass limit assumed. The inherent high temperature capability of the reactors considered removes the reactor as a limiting condition on system performance.

  8. 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.

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

    SciTech Connect

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

    1995-01-01

    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 (TFEs) to radiator heat pipes.

  10. 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.

  11. Physics of hydride fueled PWR

    NASA Astrophysics Data System (ADS)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  12. 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.

  13. Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

    SciTech Connect

    Weber, J.W.

    1994-09-28

    The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

  14. 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.

  15. 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.

  16. 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)

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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)

  4. 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.

  5. 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.

  6. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....

  7. Fast Reactor Alternative Studies: Effects of Transuranic Groupings on Metal and Oxide Sodium Fast Reactor Designs

    SciTech Connect

    R. Ferrer; M. Asgari; S. Bays; B. Forget

    2007-09-01

    A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design with a conversion ratio (CR) of 0.50 was selected in this study to perform perturbations on the external feed coming from Light Water Reactor Spent Nuclear Fuel (LWR SNF) and separation groupings in the reprocessing scheme. A secondary SFR design with a higher conversion ratio (CR=0.75) was also analyzed as a possible alternative, although no perturbations were applied to this model.

  8. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  9. 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.

  10. 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.

  11. 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.

  12. Core cooling under accident conditions at the high-flux beam reactor

    SciTech Connect

    Tichler, P.; Cheng, L. ); Fauske, H. )

    1991-01-01

    The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.

  13. CHARACTERIZATION OF THE LOCAL TITANIUM ENVIRONMENT IN DOPED SODIUM ALUMINUM HYDRIDE USING X-RAY ADSORPTION SPECTROSCOPY.

    SciTech Connect

    GRAETZ, J.; IGNATOV, A. YU; TYSON, T.A.; REILLY, J.J.; JOHNSON, J.

    2004-11-30

    Ti K-edge x-ray absorption spectroscopy was used to explore the local titanium environment and valence in 2-4 mol% Ti-doped sodium alanate. An estimate of the oxidation state of the dopant, based upon known standards, revealed a zero-valent titanium atom. An analysis of the near-edge and extended fine structures indicates that the Ti does not enter substitutional or interstitial sites in the NaAlH{sub 4} lattice. Rather, the Ti is located on/near the surface and is coordinated by 10.2 {+-} 1 aluminum atoms with an interatomic distance of 2.82 {+-} 0.01 {angstrom}, similar to that of TiAl{sub 3}. The Fourier transformed EXAFS spectra reveals a lack of long-range order around the Ti dopant indicating that the Ti forms nano-clusters of TiAl{sub 3}. The similarity of the spectra in the hydrided and dehydrided samples suggests that the local Ti environment is nearly invariant during hydrogen cycling.

  14. 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)

  15. 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.

  16. 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.

  17. 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.

  18. 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.}

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Pour stream breakup of molten IFR (Integral Fast Reactor) metal fuel in sodium

    SciTech Connect

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

    1988-01-01

    Tests have been conducted in which the breakup behavior of kilogram quantities of molten uranium, uranium-zirconium alloy, and uranium-iron alloy pour streams in 600C sodium was studied. A sodium depth of less than 0.3 m was required for hydrodynamic breakup and freezing of 25-mm pour streams of uranium and uranium-zirconium alloy with up to 400C melt superheat. The breakup material was primarily in the form of filaments and sheets with a settled bed voidage on the order of 0.9. The uranium-iron alloy with 800C melt superheat exhibited similar behavior except a sodium depth somewhat greater than 0.3 m was required for breakup and freezing of the particles.

  7. 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.

  8. 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.

  9. Specific power of liquid-metal-cooled reactors

    SciTech Connect

    Dobranich, D.

    1987-10-01

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs.

  10. 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)

  11. 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

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

    NASA Astrophysics Data System (ADS)

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

    2010-12-01

    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

  13. 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

  14. Thermal storage for solar cooling using paired ammoniated salt reactors

    NASA Astrophysics Data System (ADS)

    1981-09-01

    The feasibility of using various solid and liquid ammoniates in heat pump-thermal storage systems for space heating and cooling was conducted. Corrosion testing of selected metallic and non-metallic specimens in the ammoniates was investigated. Results of the corrosion testing showed that problems exist with manganese and magnesium chloride ammoniates, except with the teflon which displayed excellent resistance in all environments. Also, all liquid ammoniates are unsuitable for use with uncoated carbon steel. Cycling of the manganese chloride between the high and low ammoniates does not affect its properties. However, the density change between the high and low ammoniates could cause packing problems in a reactor which constrains the salt volume. Subscale tests with solid ammoniates indicated that the heat transfer coefficient in a fixed bed reactor is low (approx. 1 Btu/h-ft(2)-OF). Therefore solid ammoniates are not practical because of the high heat exchanger cost requirement.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Testability of a heat pipe cooled thermionic reactor

    NASA Astrophysics Data System (ADS)

    Durand, Richard E.; Harlan Horner, M.

    1992-01-01

    As part of the Air Force Phillips Laboratory thermionics program, Rocketdyne performed a design study for an in-core thermionic fuel element (TFE) heat pipe cooled reactor power system. This effort involved a testability evaluation that was performed starting with testing of individual components, followed by testing at various stages of fabrication, and concluding with full system acceptance and qualification testing. It was determined that the system could be thoroughly tested to ensure a high probability of successful operation in space after launch.

  20. 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.

  1. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  2. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  3. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  4. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  5. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  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. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    NASA Astrophysics Data System (ADS)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  16. 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)

  17. 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.

  18. Development of a New Thermochemical and Electrolytic Hybrid Hydrogen Production System for Sodium Cooled FBR

    NASA Astrophysics Data System (ADS)

    Nakagiri, Toshio; Kase, Takeshi; Kato, Shoichi; Aoto, Kazumi

    A new thermo-chemical and electrolytic hybrid hydrogen production system in lower temperature range is newly proposed by the Japan Nuclear Cycle Development Institute (JAEA) to realize the hydrogen production from water by using the heat generation of sodium cooled Fast Breeder Reactor (FBR). The system is based on sulfuric acid (H2SO4) synthesis and decomposition process developed earlier (Westinghouse process), and sulfur trioxide (SO3) decomposition process is facilitated by electrolysis with ionic oxygen conductive solid electrolyte to reduce the operation temperature 200-300°C lower than Westinghouse process. SO3 decomposition with the voltage lower than 0.5V was confirmed in the temperature range of 500 to 600°C and theoretical thermal efficiency of the system evaluated based on chemical reactions was within the range of 35% to 55% under the influence of H2SO4 concentration and heat recovery. Furthermore, hydrogen production experiments to substantiate the whole process were performed. Stable hydrogen and oxygen production were observed in the experiments, and maximum duration of the experiments was about 5 hours.

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

    DOE PAGES

    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

  20. 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.

  1. Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    SciTech Connect

    Silady, F.A.; Millunzi, A.C.

    1989-08-01

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab.

  2. Hydration status and fluid and sodium balance in elite Canadian junior women's soccer players in a cool environment.

    PubMed

    Gibson, Jennifer C; Stuart-Hill, Lynneth A; Pethick, Wendy; Gaul, Catherine A

    2012-10-01

    Dehydration can impair mental and on-field performance in soccer athletes; however, there is little data available from the female adolescent player. There is a lack of research investigating fluid and electrolyte losses in cool temperatures. Therefore, the purpose of this study was to investigate the pretraining hydration status, fluid balance, and sweat sodium loss in 34 female Canadian junior elite soccer athletes (mean age ± SD, 15.7 ± 0.7 years) in a cool environment. Data were collected during two 90 min on-field training sessions (9.8 ± 3.3 °C, 63% ± 12% relative humidity). Prepractice urine specific gravity (USG), sweat loss (pre- and post-training body mass), and sweat sodium concentration (regional sweat patch method) were measured at each session. Paired t tests were used to identify significant differences between training sessions and Pearson's product moment correlation analysis was used to assess any relationships between selected variables (p ≤ 0.05). We found that 45% of players presented to practice in a hypohydrated state (USG > 1.020). Mean percent body mass loss was 0.84% ± 0.07% and sweat loss was 0.69 ± 0.54 L. Although available during each training session, fluid intake was low (63.6% of players consumed <250 mL). Mean sweat sodium concentration was 48 ± 12 mmol·L⁻¹. Despite low sweat and moderate sodium losses, players did not drink enough to avoid mild fluid and sodium deficits during training. The findings from this study highlights the individual variations that occur in hydration management in athletes and thus the need for personalized hydration guidelines.

  3. 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

  4. 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

  5. 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.

  6. Environmental aspects of MHTGR (Modular High-Temperature Gas-Cooled Reactor) operation

    SciTech Connect

    Neylan, A.J.; Dilling, D.A.; Cardito, J.M.

    1988-09-01

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept being developed under a cooperative program involving the US Government, the utilities and the nuclear industry. This plant design utilizes basic High Temperature Gas-Cooled Reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The MHTGR design approach leading to exceptional safety performance also leads to plant operation which is characterized by extremely low radiological emissions even for very low probability accidents. Coated fuel particles retain radionuclides within the fuel, thus minimizing material contamination and personnel exposure. The objective of this paper is to characterize radioactive effluents expected from the normal operation of an MHTGR. In addition, other nonradioactive effluents associated with a power generating facility are discussed. Nuclear power plants produce radioactive effluents during normal operation in gaseous, liquid and solid forms. Principal sources of radioactive waste within the MHTGR are identified. The manner in which it is planned to treat these wastes is described. Like other reactors, the MHTGR produces nonradioactive effluents associated with heat generation and chemical usage. However, due to the MHTGR's higher efficiency, water usage requirements and chemical discharges for the MHTGR are minimized relative to other types of nuclear power plants. Based upon prior operating HTGR experience and analysis, effluents are quantified in terms of radioactivity levels and/or emission volume. Results, quantified within the paper, demonstrate that effluents from the MHTGR are well below regulatory limits and that the MHTGR has a minimal impact upon the public and the environment. 14 refs., 2 figs., 4 tabs.

  7. 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)

  8. 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.

  9. Conceptual Design of the Liquid Hydrogen Moderator Cooling Circuit for the European Spallation Source

    NASA Astrophysics Data System (ADS)

    Klaus, M.; Haberstroh, Ch.; Quack, H.; Beßler, Y.; Butzek, M.

    The European Spallation Sourcein Lund, Sweden, will be a 5 MW beam power neutron spallation research center. As subsystem of the target station the moderators play a vital role by slowing down high energy neutrons set free during the spallation process. To provide maximum neutron flux intensities with high availability for scattering experiments a conceptual liquid hydrogen moderator cooling circulation design proposal was developed. Supercritical hydrogen at 17 K will be utilized to absorb energy of the incoming neutrons in two parallel moderator vessels. A helium refrigerator provides the necessary cooling capacity by implementing an additional helium expansion turbine downstream the refrigerator coldbox. Strategies for the mitigation of pressure fluctuations due to beam trips are being presented. Solutions in form of electrical heaters and an accumulator or an expansion vessel are discussed. Different supercritical hydrogen circulator implementation scenarios are being matched to indicate the most reliable setup. For an efficient moderation process parahydrogen concentrations higher than 99% have to be guaranteed at the moderator inlet. Due to potential conversion of parahydrogen to orthohydrogen via irradiation processes the implementation of an ortho-parahydrogen catalyst bed is being evaluated. Methods for a continuous measurement of the apparent parahydrogen concentration at the moderator in- and outlet will be introduced. The arrangement and interaction of the components will be detailed in the paper.

  10. 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.

  11. 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.

  12. 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.

  13. 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)

  14. Cell configuration effect on feasibility of water cooled thorium breeder reactor

    SciTech Connect

    Permana, S.; Takaki, N.; Sekimoto, H.

    2006-07-01

    As a fuel candidate, thorium cycle shows some advantages such as good breeding capability, higher performance of burn-up and from proliferation point of view, thorium is more proliferation resistant. The shipping-port reactor and molten salt breeder reactor showed that breeding is possible with thorium in a thermal spectrum. Breeding is made possible by the high value of neutron regeneration ratio {eta} for {sup 233}U in thermal energy region. In the present study, feasibility of water cooled thorium breeder reactor is investigated. A calculation method by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2002 code have been performed. The reactor is fueled by {sup 233}U-Th Oxide and it has used the light water coolant and zircaloy-4 as moderator and cladding, respectively. The key properties such as flux, enrichment, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The different pin cell types have been investigated in order to analyze the effect of different fuel pin diameter. The results show the feasibility of breeding for different fuel pin cell types. The required {sup 233}U enrichment is about 2% - 9% as initial fissile loading. 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. The thicker pin cell shows wider feasible areas and requires lower enrichment than thinner pin cell. It means that thicker fuel pin diameter obtains better performances for breeding and reducing the fissile material utilization. (authors)

  15. 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)

  16. Molecular Dynamics Simulations of Aqueous and Confined Systems Relevant to the Supercritical Water Cooled Nuclear Reactor

    NASA Astrophysics Data System (ADS)

    Kallikragas, Dimitrios Theofanis

    Supercritical water (SCW) is the intended heat transfer fluid and potential neutron moderator in the proposed GEN-IV Supercritical Water Cooled Reactor (SCWR). The oxidative environment poses challenges in choosing appropriate design materials, and the behaviour of SCW within crevices of the passivation layer is needed for developing a corrosion control strategy to minimize corrosion. Molecular Dynamics simulations have been employed to obtain diffusion coefficients, coordination number and surface density characteristics, of water and chloride in nanometer-spaced iron hydroxide surfaces. Diffusion models for hydrazine are evaluated along with hydration data. Results demonstrate that water is more likely to accumulate on the surface at low density conditions. The effect of confinement on the water structure diminishes as the gap size increases. The diffusion coefficient of chloride decreases with larger surface spacing. Clustering of water at the surface implies that the SCWR will be most susceptible to pitting corrosion and stress corrosion cracking.

  17. 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.

  18. 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)

  19. 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.

  20. 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)

  1. 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

  2. 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.

  3. Role of small lead-cooled fast reactors for international deployment in worldwide sustainable nuclear energy supply.

    SciTech Connect

    Sienicki, J. J.; Wade, D. C.; Moisseytsev, A.; Nuclear Engineering Division

    2008-01-01

    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) has been under ongoing development as part of the US advanced nuclear energy systems programs. Meeting future worldwide projected energy demands during this century (e.g., 1000 to 2000 GWe by 2050) in a sustainable manner while maintaining CO2 emissions at or below today's level will require massive deployments of nuclear reactors in non-fuel cycle states as well as fuel cycle states. The projected energy demands of non-fuel cycle states will not be met solely through the deployment of Light Water Reactors (LWRs) in those states without using up the world's resources of fissile material (e.g., known plus speculative virgin uranium resources = 15 million tonnes). The present U.S. policy is focused upon domestic deployment of large-scale LWRs and sodium-cooled fast spectrum Advanced Burner Reactors (ABRs) working in a symbiotic relationship that burns existing fissile material while destroying the actinides which are generated. Other major nuclear nations are carrying out the development and deployment of SFR breeders as witness the planning for SFR breeder deployments in France, Japan, China, India, and Russia. Small (less that 300 MWe) and medium (300 to 700 MWe) size reactors are better suited to the growing economies and infrastructures of many non-fuel cycle states and developing nations. For those deployments, fast reactor converters which are fissile self-sufficient by creating as much fissile material as they consume are preferred to breeders that create more fissile material than they consume. Thus

  4. 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.

  5. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    SciTech Connect

    Caira, M.; Gramiccia, L.; Naviglio, A.; Sorabella, L.; Bandini, G.

    1996-07-01

    The MARS nuclear plant is a 600 MWth PWR with completely passive core safeguards. The most relevant innovative safety system is the Emergency Core Cooling System (ECCS), which is based on natural circulation, and on a passive-type activation that follows a core flow decrease, whatever was the cause (only one component, 400% redundant, is not static). The main thermal hydraulic transients occurring as a consequence of design basis accidents for the MARS plant were presented at the ICONE 3 Conference. Those transients were analyzed in the first stage, with the aim at pointing out the capability of the innovative ECCS to intervene. So, they included only a short-time analysis (extended for a few hundreds of seconds) and the well known RELAP 5 computer program was used for this purpose. In the present paper, the long-term analyses (extended for several thousands of seconds) of the same transients are shown. These analyses confirmed that the performance of the Emergency Core Cooling System of the MARS reactor is guaranteed also in long-term scenarios.

  6. 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

  7. 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.

  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. Conceptual studies for pressurised water reactor cores employing plutonium erbium zirconium oxide inert matrix fuel assemblies

    NASA Astrophysics Data System (ADS)

    Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

    1999-08-01

    The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor (PWR), the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2-Er 2O 3-ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to `real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2-fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies.

  10. 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.

  11. 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.

  12. 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.

  13. 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)

  14. 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 .

  15. Sodium technology activities at HEDL in support of fast reactor development and the FFTF

    SciTech Connect

    Atwood, J.M.

    1984-02-27

    Activities of the Hanford Engineering Development Laboratory are presented. A brief description of FFTF and some highlights of reactor operations are reviewed. The sodium technology work at HEDL is summarized by discussing several facets of the program and their tie-ins to breeder reactor development.

  16. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    SciTech Connect

    Pudelek, R. E.; Gilbert, W. C.

    2002-02-26

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste

  17. 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

  18. 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)

  19. 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.

  20. 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.

  1. Identification of galactose as the immunodominant sugar of leishmanial excreted factor and subsequent labeling with galactose oxidase and sodium boro[3H]hydride.

    PubMed

    Slutzky, G M; Greenblatt, C L

    1982-07-01

    Inhibition by low-molecular-weight sugars of precipitin line formation between a polysaccharide (EF) excreted by Leishmania tropica subsp. major, Leishmania enriettii, and rabbit antileishmanial antibodies on double gel diffusion plates revealed that galactose residues, possibly as components of lactosyl groups, were the critical immunodominant sugars mediating antibody recognition of EF. The galactose residues of the EF of L. tropica subsp. major were specifically labeled with tritium via galactose oxidase and sodium boro[3H]hydride. The radioactive EF had an apparent molecular weight of about 85,000 on sodium dodecyl sulfate-polyacrylamide gels and was precipitated by antileishmanial antibodies as well as Ricinus communis lectins I and II (galactose specific). Lectins specific for glucose-mannose residues, fucose, N-acetylglucosamine, and N-acetylgalactosamine did not precipitate the labeled EF. Treatment of [3H]EF with proteolytic (trypsin, papain, protease) or glycosidic (alpha-amylase, beta-galactosidase) enzymes had no effect on either the electrophoretic pattern of the material or on its recognition by antileishmanial antibodies or R. communis lectin. This resistance to enzyme activity suggests that EF may be a useful marker for the presence of the parasite in vivo if it can be detected in minute quantities. PMID:6179874

  2. 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)

  3. A primary sodium pump gene of the moderate halophile Halobacillus dabanensis exhibits secondary antiporter properties.

    PubMed

    Yang, Lifu; Jiang, Juquan; Zhang, Bo; Zhao, Baisuo; Wang, Lei; Yang, Su Sheng

    2006-07-28

    The primary sodium pump has been proved to be involved in Na(+) extrusion of bacteria. In our present study, a novel gene encoding a putative primary sodium pump was cloned from chromosomal DNA of moderate halophile Halobacillus dabanensis D-8 by functional complementation, which expression resulted in the growth of antiporter-deficient Escherichia coli strain KNabc in the presence of 0.2 M NaCl. The gene was sequenced and designated nap. The deduced amino acid sequence of Nap has 56% identity to NADH dehydrogenase of Bacillus cereus and 55% to NADH oxidase of Bacillus halodurans C-125. E. coli KNabc carrying nap exhibited resistance to uncoupler CCCP (carbonyl-cyanide m-chlorophenylhydrazone). Everted membrane vesicles prepared from E. coli KNabc carrying nap exhibited secondary Na(+)/H(+) antiporter activity, and nap also supported the growth of respiratory-deficient E. coli ANN0222 lacking NADH dehydrogenase. Based on these results, we proposed that Nap possessed both characteristics of secondary Na(+)/H(+) antiporter and primary sodium pump.

  4. 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.

  5. 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.

  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. 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.

  8. Spectrophotometric determination of zirconium with xylenol orange

    SciTech Connect

    Antepenko, R J

    1982-05-14

    High purity hydride forming metal films are used as hydrogen isotope occluders and function as electrodes in neutron generator tubes. This use of zirconium occluder films requires reliable analytical methods for routine determination of the zirconium film weight in a production environment. In this study, a spectrophotometric method was evaluated for the determination of zirconium films. The method is based upon the formation of a highly colored zirconium complex with xylenol orange in a dilute perchloric acid medium. Dilute hydrofluoric acid is used in this procedure to selectively dissolve the zirconium film off the substrate. A perchloric acid fuming step is used to remove hydrofluoric acid from the solution. The zirconium solutions are depolymerized before complex formation by heating in 2 N perchloric acid. The zirconium complex exhibits a maximum absorbance in 0.2 to 0.3 M perchloric acid at a wavelength of 531 nanometers. Beer's law is obeyed for zirconium concentrations through 2.1 parts per million. Molybdenum, at concentrations equal to zirconium, does not interfere with the xylenol orange method.

  9. 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.

  10. 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.

  11. 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.

  12. Scaling Analysis for the Direct Reactor Auxillary Cooling System For AHTRS

    SciTech Connect

    Lv, Q. NMN; Wang, X. NMN; Sun, X NMN; Christensen, R. N.; Blue, T. E.; Yoder Jr, Graydon L; Wilson, Dane F; Subharwall, Piyush; Adams, I.

    2013-01-01

    The Direct Reactor Auxiliary Cooling System (DRACS) is a passive heat removal system proposed for the Advanced High-Temperature Reactor (AHTR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops relying completely on buoyancy as the driving force. In the DRACS, two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX) are used to couple these loops. In addition, a fluidic diode is employed to minimize the parasitic flow during normal operation of the reactor and to activate the DRACS in accidents. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for AHTRs built and tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of the 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 straightforwardly 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 also been developed, which consists of the core scaling and loop scaling. The consistence between the core and loop scaling is examined through the reference volume ratio, which can be obtained from the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a design of the scaled-down high-temperature DRACS test facility (HTDF).

  13. 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.

  14. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    SciTech Connect

    Clayton, J C

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated.

  15. 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.

  16. 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.

  17. 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.

  18. Operating Temperatures of a Sodium-Cooled Exhaust Valve as Measured by a Thermocouple

    NASA Technical Reports Server (NTRS)

    Sanders, J C; Wilsted, H D; Mulcahy, B A

    1943-01-01

    Report presents the results of a thermocouple installed in the crown of a sodium-cooled exhaust valve. The valve was tested in an air-cooled engine cylinder and valve temperatures under various engine operating conditions were determined. A temperature of 1337 degrees F. was observed at a fuel-air ratio of 0.064, a brake mean effective pressure of 179 pounds per square inch, and an engine speed of 2000 r.p.m. Fuel-air ratio was found to have a large influence on valve temperature, but cooling-air pressure and variation in spark advance had little effect. An increase in engine power by change of speed or mean effective pressure increased the valve temperature. It was found that the temperature of the rear-spark-plug bushing was not a satisfactory indication of the temperature of the exhaust valve.

  19. Operating Temperatures of a Sodium-Cooled Exhaust Valve as Measured by a Thermocouple

    NASA Technical Reports Server (NTRS)

    Sanders, J. C.; Wilsted, H. D.; Mulcahy, B. A.

    1943-01-01

    A thermocouple was installed in the crown of a sodium-cooled exhaust valve. The valve was then tested in an air-cooled engine cylinder and valve temperatures under various engine operating conditions were determined. A temperature of 1337 F was observed at a fuel-air ratio of 0.064, a brake mean effective pressure of 179 pounds per square inch, and an engine speed of 2000 rpm. Fuel-air ratio was found to have a large influence on valve temperature, but cooling-air pressure and variation in spark advance had little effect. An increase in engine power by change of speed or mean effective pressure increased the valve temperature. It was found that the temperature of the rear spark-plug bushing was not a satisfactory indication of the temperature of the exhaust valve.

  20. 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.

  1. 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.

  2. 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

  3. Electrochemical process for zirconium alloy recycling

    SciTech Connect

    Snyder, T.S.; Stoltz, R.A.; Zuckerbrod, D.

    1990-05-08

    This patent describes a method of separating nickel from zirconium for recycling nickel-containing zirconium alloy. It comprises: placing the nickel-containing zirconium in a molten alt bath at 500{degrees} to about 722{degrees} C. with the molten salt in the molten salt bath consisting essentially of a mixture of 0 to about 46.5 mole % lithium fluoride, about 11.5 to about 40 mole % sodium fluoride and potassium fluoride to produce a molten salt bath containing dissolved zirconium and dissolved nickel; electrochemically plating the nickel from the molten salt bath at a voltage sufficient to plate nickel but less than the voltage to plate zirconium to provide an essentially nickel-free molten salt bath; and electrochemically plating the zirconium from the essentially nickel-free molten salt bath to provide an essentially nickel-free zirconium.

  4. 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.

  5. 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.

  6. 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.

  7. Cooling arrangement for the locking device in pressure gasification reactors

    SciTech Connect

    Jaschke, P.; Monch, E.; Weber, R.

    1981-06-02

    A cooling arrangement particularly for cooling a locking device arranged in a fuel feed opening of pressure gasification generators comprises a lock seat connected to the generator and a lock cone adapted for moving into and out of engagement with the lock seat and provided with a supply line to provide a cooling medium to the lock cone and a discharge line through which the used medium is drained off the lock cone. The lock cone is formed as a hollow body with a cooling chamber in its interior and is provided with guide means extending towards the lowermost area of the lock cone to direct the cooling medium into contact with the mostly overheated lowermost area of the lock cone positioned in a close relationship with the pressure gasification generator.

  8. 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

  9. 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)

  10. Microstructural studies and crystallographic orientation of different zones and δ-hydrides in resistance welded Zircaloy-4 sheets

    NASA Astrophysics Data System (ADS)

    Kiran Kumar, N. A. P.; Szpunar, Jerzy. A.; He, Zhang

    2011-07-01

    The cold worked stress relieved (CWSR) Zircaloy-4 sheet used as endplate in nuclear fuel bundle is resistance welded with an endcap in argon environment. Later the welded sample is hydrided in a gaseous atmosphere at 400 °C. Optical microscopy (OM), electron backscatter diffraction (EBSD) and X-ray diffraction (XRD) were used to examine the morphology and crystal orientation of the hydrides. The microstructural changes in different areas of the weld zone, heat affected zone (HAZ) and the as-received zone were analyzed using EBSD technique. Optical examination showed complete random morphological orientation of hydrides and predominantly basket-weave structure in the weld zone, with very few colonies of parallel plate structures. Variant selection for α-phase formation inside prior β-grains was identified at the weld centre. As we move from the weld centre to the as-received zone, the variant selection is found to be less probable. The δ-hydride platelets at the weld zone were always found to be growing perpendicular to the α-colonies having angular difference of 60-63° and follow (0 0 0 1) α-Zr//{1 1 1}δ-ZrH 1.5 orientation relationship with the zirconium matrix. Proposed description of complex distribution of hydrides and alloy microstructure at the weld and heat affected zone will contribute to a better understanding of mechanisms of failure of fuel cladding in various types of nuclear reactors.

  11. 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

  12. 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.

  13. Overview of environmental control aspects for the gas-cooled fast reactor

    SciTech Connect

    Nolan, A.M.

    1981-05-01

    Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included.

  14. 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.

  15. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative)

    NASA Astrophysics Data System (ADS)

    Marshall, A. C.

    1989-08-01

    A review was completed of multimegawatt gas cooled reactor concepts proposed for SDI applications. The 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 was 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.

  16. 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.

  17. 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.

  18. Investigation of a para-ortho hydrogen reactor for application to spacecraft sensor cooling

    NASA Astrophysics Data System (ADS)

    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.

  19. 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

  20. 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.

  1. 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.

  2. Interstellar Hydrides

    NASA Astrophysics Data System (ADS)

    Gerin, Maryvonne; Neufeld, David A.; Goicoechea, Javier R.

    2016-09-01

    Interstellar hydrides—that is, molecules containing a single heavy element atom with one or more hydrogen atoms—were among the first molecules detected outside the solar system. They lie at the root of interstellar chemistry, being among the first species to form in initially atomic gas, along with molecular hydrogen and its associated ions. Because the chemical pathways leading to the formation of interstellar hydrides are relatively simple, the analysis of the observed abundances is relatively straightforward and provides key information about the environments where hydrides are found. Recent years have seen rapid progress in our understanding of interstellar hydrides, thanks largely to FIR and submillimeter observations performed with the Herschel Space Observatory. In this review, we discuss observations of interstellar hydrides, along with the advanced modeling approaches that have been used to interpret them and the unique information that has thereby been obtained.

  3. 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.

  4. Engineering design of a liquid metal cooled self-pumped limiter for a tokamak reactor

    SciTech Connect

    Brooks, J.N.; Cha, Y.; Hassanein, A.; Majumdar, S.; Mattas, R.F.; Smith, D.L.

    1987-10-01

    A lithium cooled self-pumped limiter has been designed as the impurity control system for the TPSS high-..beta.. power reactor conceptual design. The limiter removes helium by trapping impinging helium ions in freshly deposited vanadium surface layers in a slot region. No hydrogen is removed and no pumps or vacuum penetrations are used, thereby eliminating penetration shielding and reducing tritium handling. The limiter is composed of a vanadium alloy structure with a 2mm tungsten cladding on the front face and leading edges for sputtering control. Up to approx.3cm of vanadium trapping material is deposited in the slot region during 5 years of operation. A key design feature is the use of a calcium oxide electrical insulator which coats the limiter coolant channels to reduce MHD pressure drops. A combination of high lithium coolant velocity, made possible by the insulator, and mid-limiter manifolding has been used to obtain acceptable material temperatures with moderately high heat fluxes (3 to 5 mw/m/sup 2/). Overall, a limiter lifetime of approx.5 years is predicted by stress and lifetime analysis. This would permit maintenance free impurity control operation between first wall/blanket replacement periods. 4 refs., 3 figs., 1 tab.

  5. Multi-stage hydride-hydrogen compressor

    NASA Astrophysics Data System (ADS)

    Golben, P. M.

    A 4-stage metal hydride/hydrogen compressor that uses low temperature hot water (75 C) as its energy source has been built and tested. The compressor utilizes a new hydride heat exchanger technique that has achieved fast cycling time (with 20 C cooling water) on the order of 1 min. This refinement substantially decreases the size, weight and cost of the unit when compared to previous hydride compressors or even conventional mechanical diaphragm compressors.

  6. 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.

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  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. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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.

  20. 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.

  1. 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)

  2. 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)

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. Transport of breeder reactor-fire-generated sodium oxide aerosols for building-wake-dominated meteorology

    SciTech Connect

    Fields, D.E.; Cooper, A.C.; Miller, C.W.

    1987-02-01

    This report describes the methodology used and results obtained in efforts to estimate the sodium aerosol concentrations at air intake ports of a liquid-metal cooled, fast-breeder nuclear reactor. An earlier version of this methodology has been previously discussed (Fields and Miller, 1985). A range of wind speeds from 2 to 10 m/s is assumed, and an effort is made to include building wake effects which, in many cases, dominate the dispersal of aerosols near buildings. For relatively small release rates, on the order of 1 to 10 kg/s, the plume rise is small and estimates of aerosol concentrations are derived using the methodology of Wilson and Britter (1982), which describes releases from surface vents. For release rates on the order of 100 kg/s much higher release velocities are expected, and plume rise is considered. An effective increase in release height is computed using the Split-H methodology with a parameterization suggested by Ramsdell (1983), and the release source strength is transformed to rooftop level. Evaluation of the acute release aerosol concentration is then based on the methodology for releases from a surface release of this transformed source strength. For a horizontal release, a methodology is developed to chart the plume path as a function of release and site meteorology parameters. Results described herein must be regarded as maximum aerosol concentrations, based on models derived from generic wind tunnel studies. More accurate and site-specific results may be obtained through wind tunnel simulations and through simulating emissions from release points other than those assumed here.

  8. 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.

  9. 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)

  10. 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.

  11. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  12. Development of a silicon calorimeter for dosimetry applications in a water-moderated reactor.

    SciTech Connect

    DePriest, Kendall Russell; King, Donald Bryan; Naranjo, Gerald E.; Luker, Spencer Michael; Keltner, Ned R.; Suo-Anttila, Ahti Jorma; Griffin, Patrick Joseph

    2005-05-01

    High fidelity active dosimetry in the mixed neutron/gamma field of a research reactor is a very complex issue. For passive dosimetry applications, the use of activation foils addresses the neutron environment while the use of low neutron response CaF{sub 2}:Mn thermoluminescent dosimeters (TLDs) addresses the gamma environment. While radiation-hardened diamond photoconducting detectors (PCD) have been developed that provide a very precise fast response (picosecond) dosimeter and can provide a time-dependent profile for the radiation environment, the mixed field response of the PCD is still uncertain and this interferes with the calibration of the PCD response. In order to address the research reactor experimenter's need for a dosimeter that reports silicon dose and dose rate at a test location during a pulsed reactor operation, a silicon calorimeter has been developed. This dosimeter can be used by itself to provide a dose in rad(Si) up to a point in a reactor pulsed operation, or, in conjunction with the diamond PCD, to provide a dose rate. This paper reports on the development, testing, and validation of this silicon calorimeter for applications in water-moderated research reactors.

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. 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.

  18. 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).

  19. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... document. You may access publicly-available information related to this action by the following methods... CONTACT section of this document. NRC's Agencywide Documents Access and Management System (ADAMS): You may... emergency core cooling systems (ECCSs) of pressurized water reactors (PWRs). This RG also describes...

  20. 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

  1. Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

    SciTech Connect

    Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

    2012-05-22

    There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  10. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  11. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  12. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  13. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  14. 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.

  15. 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.

  16. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 2A: Accident model document, appendix

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The detailed abort sequence trees for the reference zirconium hydride (ZrH) reactor power module that have been generated for each phase of the reference Space Base program mission are presented. The trees are graphical representations of causal sequences. Each tree begins with the phase identification and the dichotomy between success and failure. The success branch shows the mission phase objective as being achieved. The failure branch is subdivided, as conditions require, into various primary initiating abort conditions.

  17. 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.

  18. New approaches in the USA for high-temperature gas-cooled reactors. Gas-Cooled Reactor Programs

    SciTech Connect

    Kasten, P.R.; Neylan, A.J.; Penfield, S.R. Jr.

    1985-08-01

    Several concepts 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 reactor power and increase the ability to use inherent features for removing heat following extreme accidents. The unit size and design of these concepts are constrained so that extreme accidents do not result in significant release of radioactivity from the reactor plant. Through the greater reliance on inherent safety features in small HTRs, it should be possible to minimize the amount of nuclear grade components required in the balance-of-plant, which could lead to an economic system. Four HTR concepts are presently being evaluated within the US Program, and these concepts are briefly summarized. A modular HTR using a steel pressure vessel, which is very similar to one of the four HTR concepts being evaluated within the US National program, is presented as an example of a specific concept to illustrate the features and performance of HTRs having a high degree of inherent safety.

  19. Passive decay heat removal system for water-cooled nuclear reactors

    DOEpatents

    Forsberg, Charles W.

    1991-01-01

    A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

  20. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    SciTech Connect

    Sweeney, F.J. ); Carroll, D.G. ); Chen, C. ); Crane, C.; Dalton, R. ); Taylor, J.R. ); Tosunoglu, S. )

    1993-01-01

    One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS.

  1. A Subcritical, Gas-Cooled Fast Transmutation Reactor with a Fusion Neutron Source

    SciTech Connect

    Stacey, W.M.; Beavers, V.L.; Casino, W.A.; Cheatham, J.R.; Friis, Z.W.; Green, R.D.; Hamilton, W.R.; Haufler, K.W.; Hutchinson, J.D.; Lackey, W.J.; Lorio, R.A.; Maddox, J.W.; Mandrekas, J.; Manzoor, A.A.; Noelke, C.A.; Oliveira, C. de; Park, M.; Tedder, D.W.; Terry, M.R.; Hoffman, E.A.

    2005-05-15

    A design is presented for a subcritical, He-cooled fast reactor, driven by a tokamak D-T fusion neutron source, for the transmutation of spent nuclear fuel (SNF). The reactor is fueled with coated transuranic (TRU) particles and is intended for the deep-burn (>90%) transmutation of the TRUs in SNF without reprocessing of the coated fuel particles. The reactor design is based on the materials, fuel, and separations technologies under near-term development in the U.S. Department of Energy (DOE) Nuclear Energy Program and on the plasma physics and fusion technologies under near-term development in the DOE Fusion Energy Sciences Program, with the objective of intermediate-term ({approx}2040) deployment. The physical and performance characteristics and research and development requirements of such a reactor are described.

  2. a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

    NASA Astrophysics Data System (ADS)

    Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

    2009-08-01

    This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

  3. [Toxicity and pharmacokinetics of zirconium oxychloride in mice and rats].

    PubMed

    Delongeas, J L; Burnel, D; Netter, P; Grignon, M; Mur, J M; Royer, R J; Grignon, G

    1983-01-01

    The experimental toxicity of zirconium compounds is examined in the Mouse (acute toxicity) and in the Rat (short time toxicity). The absorption, the distribution and the elimination of zirconium are evaluated by zirconium cation assay in some biological fluids and tissues. After a single oral dose, zirconium oxyd is not toxic, zirconium oxychlorure slightly toxic and zirconium chlorure moderately toxic. At certain concentrations, cerebral and pulmonary disorders are observed, particularly with zirconium chlorure. In considering molar toxicity, the studied zirconium compounds are more toxic than certain aluminium salts mentioned in the literature. The zirconium oxychlorure doesn't influence the growth curve after iterative administrations (0.23 g zirconium/kg/day). Only a weak fraction of administered zirconium is absorbed and is electively fixed in the ovaries, in a lesser degree in the lung and the bone. In the ovary the zirconium induces vascular variation (hypervascularization) which appear one month after the end of the treatment. The absorbed zirconium is eliminated by the urinary tract. The fecal elimination can be essentially explained by an important quantity of non absorbed zirconium. PMID:6672464

  4. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    NASA Astrophysics Data System (ADS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S∧4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S∧4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively.

  5. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    SciTech Connect

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-20

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respect0011ive.

  6. Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)

    SciTech Connect

    Massie, M.; Forsberg, C.; Forget, B.; Hu, L. W.

    2012-07-01

    A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

  7. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

    SciTech Connect

    Forsberg, C. W.; Terrani, Kurt A; Snead, Lance Lewis; Katoh, Yutai

    2012-01-01

    The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of {sup 7}LiF and BeF{sub 2} (FLiBe) possessing a boiling point above 1300 C and the figure of merit {rho}C{sub p} (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

  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. Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

    SciTech Connect

    Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael; Walker, Matthew

    2016-01-01

    This report outlines the thermodynamics of a supercritical carbon dioxide (sCO2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

  10. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    SciTech Connect

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  11. A 50-100 kWe gas-cooled reactor for use on Mars.

    SciTech Connect

    Peters, Curtis D.

    2006-04-01

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

  12. Modular high-temperature gas-cooled reactor core heatup accident simulations

    SciTech Connect

    Ball, S.J.; Conklin, J.C.

    1989-01-01

    The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs.

  13. Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

    SciTech Connect

    Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.; Phillips, Jesse; Kalinich, Donald A.; Osborn, Douglas M.; Peko, Damian

    2013-11-01

    Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

  14. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    SciTech Connect

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

  15. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    SciTech Connect

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  16. Performance optimization considerations for thermionic fuel elements in a heat pipe cooled thermionic reactor

    NASA Astrophysics Data System (ADS)

    Bellis, Elizabeth A.

    1992-01-01

    A heat pipe-cooled, in-core thermionic (HPTI) reactor design has been proposed in support of the Air Force Thermionic Space Nuclear Power Program. As part of this design, the performance of the power conversion system has been characterized. This paper focuses on the performance optimization studies carried out of a thermionic fuel element (TFE) which will be used in a reactor design capable of producing 40 kWe over a 10 year operating life. The technical approach to the optimization studies closely couples converter lifetime constraints with converter performance to produce the best possible design choice.

  17. Modular high temperature gas-cooled reactor plant design duty cycle. Revision 3

    SciTech Connect

    Chan, T.

    1989-12-31

    This document defines the Plant Design Duty Cycle (PCDC) for the Modular High Temperature Gas-cooled Reactor (MHTGR). The duty cycle is a set of events and their design number of occurrences over the life of the plant for which the MHTGR plant shall be designed to ensure that the plant meets all the top-level requirements. The duty cycle is representative of the types of events to be expected in multiple reactor module-turbine plant configurations of the MHTGR. A synopsis of each PDDC event is presented to provide an overview of the plant response and consequence. 8 refs., 1 fig., 4 tabs.

  18. Fuel performance models for high-temperature gas-cooled reactor core design

    SciTech Connect

    Stansfield, O.M.; Simon, W.A.; Baxter, A.M.

    1983-09-01

    Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10/sup -4/ fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience.

  19. Moderation of dietary sodium potentiates the renal and cardiovascular protective effects of angiotensin receptor blockers.

    PubMed

    Lambers Heerspink, Hiddo J; Holtkamp, Frank A; Parving, Hans-Henrik; Navis, Gerjan J; Lewis, Julia B; Ritz, Eberhard; de Graeff, Pieter A; de Zeeuw, Dick

    2012-08-01

    Dietary sodium restriction has been shown to enhance the short-term response of blood pressure and albuminuria to angiotensin receptor blockers (ARBs). Whether this also enhances the long-term renal and cardiovascular protective effects of ARBs is unknown. Here we conducted a post-hoc analysis of the RENAAL and IDNT trials to test this in patients with type 2 diabetic nephropathy randomized to ARB or non-renin-angiotensin-aldosterone system (non-RAASi)-based antihypertensive therapy. Treatment effects on renal and cardiovascular outcomes were compared in subgroups based on dietary sodium intake during treatment, measured as the 24-h urinary sodium/creatinine ratio of 1177 patients with available 24-h urinary sodium measurements. ARB compared to non-RAASi-based therapy produced the greatest long-term effects on renal and cardiovascular events in the lowest tertile of sodium intake. Compared to non-RAASi, the trend in risk for renal events was significantly reduced by 43%, not changed, or increased by 37% for each tertile of increased sodium intake, respectively. The trend for cardiovascular events was significantly reduced by 37%, increased by 2% and 25%, respectively. Thus, treatment effects of ARB compared with non-RAASi-based therapy on renal and cardiovascular outcomes were greater in patients with type 2 diabetic nephropathy with lower than higher dietary sodium intake. This underscores the avoidance of excessive sodium intake, particularly in type 2 diabetic patients receiving ARB therapy.

  20. 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.