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Sample records for fission fuels

  1. Fission-product retention in HTGR fuels

    SciTech Connect

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed.

  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. A fission gas release correlation for uranium nitride fuel pins

    NASA Technical Reports Server (NTRS)

    Weinstein, M. B.; Davison, H. W.

    1973-01-01

    A model was developed to predict fission gas releases from UN fuel pins clad with various materials. The model was correlated with total release data obtained by different experimentors, over a range of fuel temperatures primarily between 1250 and 1660 K, and fuel burnups up to 4.6 percent. In the model, fission gas is transported by diffusion mechanisms to the grain boundaries where the volume grows and eventually interconnects with the outside surface of the fuel. The within grain diffusion coefficients are found from fission gas release rate data obtained using a sweep gas facility.

  4. Chemical state of fission products in irradiated uranium carbide fuel

    NASA Astrophysics Data System (ADS)

    Arai, Yasuo; Iwai, Takashi; Ohmichi, Toshihiko

    1987-12-01

    The chemical state of fission products in irradiated uranium carbide fuel has been estimated by equilibrium calculation using the SOLGASMIX-PV program. Solid state fission products are distributed to the fuel matrix, ternary compounds, carbides of fission products and intermetallic compounds among the condensed phases appearing in the irradiated uranium carbide fuel. The chemical forms are influenced by burnup as well as stoichiometry of the fuel. The results of the present study almost agree with the experimental ones reported for burnup simulated carbides.

  5. (Fuel, fission product, and graphite technology)

    SciTech Connect

    Stansfield, O.M.

    1990-07-25

    Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

  6. Fusion-Fission Hybrid for Fissile Fuel Production without Processing

    SciTech Connect

    Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

    2012-01-02

    Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in the critical reactors

  7. AFC-1 Fuel Rodlet Fission Power Deposition Validation in ATR

    SciTech Connect

    G. S. Chang; M. A. Lillo; D. J. Utterbeck

    2008-11-01

    One of the viable options of long-term geological disposal of the nuclear power reactors generated spent fuel is to extract plutonium, the minor actinides (MA) and potentially long-lived fission products from the spent fuel and transmute them into short-lived or stable radionuclides in an appropriate reactor for the reduction of the radiological toxicity of the nuclear waste stream. An important component of that technology will be a non-fertile / low-fertile actinide transmutation fuel form containing the plutonium, neptunium, americium (and possibly curium) isotopes to be transmuted. Such advanced fuel forms, especially ones enriched in the long-life minor actinide (LLMA) elements (i.e., Np, Am, Cm), have minimal irradiation performance data available from which to establish a transmutation fuel form design. Recognizing these needs, an Advanced Fuel Cycle test series-1 (AFC-1) irradiation test on a variety of candidate fuel forms is now being conducted in Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The first advanced fuel experiment (AFC-1) has been finalized and the test assembly analyzed for insertion and irradiation in ATR. The ATR core consists of a serpentine and rotationally symmetric fuel assembly about the z-axis of the core center. The plan view of the ATR core configuration is shown in Fig. 5, in Ref. 1. A cadmium filter with a 0.178 cm (0.045") thickness and 121.5 cm (48") in length, is currently used in the actinide-fuel capsule design for the East Flux Trap (EFT) position in ATR, to depress the linear heat generation rate (LHGR) lower than the project’s 330 W/cm limit for the experimental fuel rodlets. The LHGR is proportional to the fission power deposited in the fuel rodlets from the neutron fissions. The fraction of the fission power generated from the neutron fission reactions deposited in the fuel rodlet is an important parameter for test assembly thermal analysis, which will be validated in this summary.

  8. Fission induced swelling of U-Mo/Al dispersion fuel

    NASA Astrophysics Data System (ADS)

    Kim, Yeon Soo; Jeong, G. Y.; Park, J. M.; Robinson, A. B.

    2015-10-01

    Fission-induced swelling of U-Mo/Al dispersion fuel meat was measured using microscopy images obtained from post-irradiation examination. The data of reduced-size plate-type test samples and rod-type test samples were employed for this work. A model to predict the meat swelling of U-Mo/Al dispersion fuel was developed. This model is composed of several submodels including a model for interaction layer (IL) growth between U-Mo and Al matrix, a model for IL thickness to IL volume conversion, a correlation for the fission-induced swelling of U-Mo alloy particles, a correlation for the fission-induced swelling of IL, and models of U-Mo and Al consumption by IL growth. The model was validated using full-size plate data that were not included in the model development.

  9. Release of fission gas during transient heating of LWR fuel

    SciTech Connect

    Gehl, S.M.

    1982-05-01

    The direct electrical heating technique was used to study fission-gas release and mechanical behavior of irradiated light-water reactor (LWR) fuels during thermal transients. An empirical correlation between fission-gas release and transient temperature history was developed for power-cooling mismatch (PCM) and anticipated transients. Gas release during the refill portion of a design-basis loss of cooling accident was estimated to be less than 1%. Fission-gas release during PCM accidents was found to be controlled by intergranular microcracking and the interlinkage of tunnels on grain edges. For high-gas-release transients, the fractional gas release was shown to be equal to the fractional coverage of grain boundaries by microcracks. Temperature calculations indicated that microcracking causes a significant decrease in the fuel thermal conductivity.

  10. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  11. Fission gas retention and axial expansion of irradiated metallic fuel

    SciTech Connect

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

    1986-05-01

    Out-of-reactor experiments utilizing direct electrical heating and infrared heating techniques were performed on irradiated metallic fuel. The results indicate accelerated expansion can occur during thermal transients and that the accelerated expansion is driven by retained fission gases. The results also demonstrate gas retention and, hence, expansion behavior is a function of axial position within the pin.

  12. Surface Power Fission System Fuels: Technology Readiness and Qualification Issues

    SciTech Connect

    Marcille, Thomas F.; Poston, David I.

    2006-07-01

    The choice of fuel form for an initial U.S space reactor for surface power applications is significant for a timely, cost effective development program through flight system delivery. While minimizing fuel system development effort and cost is key to achieving a successful practical application, performance and reliability of that fuel system must be predictably understood to insure that the integrated mission requirements are achieved. Performance and reliability predictions depend on understanding the fundamental irradiated material properties for any fuel system (fuel meat plus clad), including thermal conductivity, thermal mechanical and irradiation-induced (constrained and unconstrained) swelling as functions of temperature, burnup and time, and fission gas release. Qualification tasks may include in-pile irradiations and associated post irradiation examinations (PIE) for thermally and spectrally representative operating environments, for intended operational fission rates. Additionally, simplification of the qualification tasks may result from the use of available database information from prior representative in-pile experiments. While available data is compelling, current manufacturing processes may differ from those employed to facilitate previous experiments, and assessing the pedigree of previous processes relative to current ones may be difficult at best. Considered fuel meats include UN, UO{sub 2}, UZrH and metal fuels. This paper will examine the state of the current qualification basis for each of these fuel materials, and assess each as appropriate for use with a low cost initial surface power fission system. A low cost system can not bear the cost of expensive material development programs, necessitated by the lack of high temperature irradiated material property data. As such, a low cost system limits clad and coolant temperatures to < 900 K. (authors)

  13. Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

    SciTech Connect

    Herrmann, Steven Douglas

    2014-05-27

    Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

  14. Release of fission tritium through Zircaloy-4 fuel cladding tubes

    NASA Astrophysics Data System (ADS)

    Andrieu, C.; Ravel, S.; Ducros, G.; Lemaignan, C.

    2005-12-01

    In order to analyse a potential fission tritium migration from the fuel to the coolant trough the cladding, two experiments of effusion and permeation types, have been performed on 3H release from Zry4 claddings. During the tests at 350 °C, the 3H released activities were measured at regular intervals. In both cases, very fast release rates have been obtained in the first few days, followed by more steady release rates. A correlation has been obtained between the 3H releases measured and the oxide formation kinetics after the initial burst. A mechanism of 3H transport is proposed based on the behaviour of the precipitates during the oxidation of Zry4. Applied to the conditions of PWR fuels, the measurements performed and the mechanisms considered lead to an insignificant contribution of fission tritium permeation to the total inventory of the tritium in the primary coolant.

  15. Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

    SciTech Connect

    Troy Unruh; Michael Reichenberger; Phillip Ugorowski

    2013-09-01

    Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

  16. Fission induced swelling and creep of U-Mo alloy fuel

    NASA Astrophysics Data System (ADS)

    Kim, Yeon Soo; Hofman, G. L.; Cheon, J. S.; Robinson, A. B.; Wachs, D. M.

    2013-06-01

    Tapering of U-Mo alloy fuel at the end of plates is attributed to lateral mass transfer by fission induced creep, by which fuel mass is relocated away from the fuel end region where fission product induced fuel swelling is in fact the highest. This mechanism permits U-Mo fuel to achieve high burnup by effectively relieving stresses at the fuel end region, where peak stresses are otherwise expected because peak fission product induced fuel swelling occurs there. ABAQUS FEA was employed to examine whether the observed phenomenon can be simulated using physical-mechanical data available in the literature. The simulation results obtained for several plates with different fuel fabrication and loading scheme showed that the measured data were able to be simulated with a reasonable creep rate coefficient. The obtained creep rate constant lies between values for pure uranium and MOX, and is greater than all other ceramic uranium fuels.

  17. Fission induced swelling and creep of U–Mo alloy fuel

    SciTech Connect

    Yeon Soo Kim; G. L. Hofman; J. S. Cheon; A. B. Robinson; D. M. Wachs

    2013-06-01

    Tapering of U–Mo alloy fuel at the end of plates is attributed to lateral mass transfer by fission induced creep, by which fuel mass is relocated away from the fuel end region where fission product induced fuel swelling is in fact the highest. This mechanism permits U–Mo fuel to achieve high burnup by effectively relieving stresses at the fuel end region, where peak stresses are otherwise expected because peak fission product induced fuel swelling occurs there. ABAQUS FEA was employed to examine whether the observed phenomenon can be simulated using physical–mechanical data available in the literature. The simulation results obtained for several plates with different fuel fabrication and loading scheme showed that the measured data were able to be simulated with a reasonable creep rate coefficient. The obtained creep rate constant lies between values for pure uranium and MOX, and is greater than all other ceramic uranium fuels.

  18. Radiation re-solution of fission gas in non-oxide nuclear fuel

    NASA Astrophysics Data System (ADS)

    Matthews, Christopher; Schwen, Daniel; Klein, Andrew C.

    2015-02-01

    Renewed interest in fast nuclear reactors is creating a need for better understanding of fission gas bubble behavior in non-oxide fuels to support very long fuel lifetimes. Collisions between fission fragments and their subsequent cascades can knock fission gas atoms out of bubbles and back into the fuel lattice. We showed that these collisions can be treated as using the so-called "homogenous" atom-by-atom re-solution theory and calculated using the Binary Collision Approximation code 3DOT. The calculations showed that there is a decrease in the re-solution parameter as bubble radius increases until about 50 nm, at which the re-solution parameter stays nearly constant. Furthermore, our model shows ion cascades created in the fuel result in many more implanted fission gas atoms than collisions directly with fission fragments. This calculated re-solution parameter can be used to find a re-solution rate for future bubble simulations.

  19. HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

    SciTech Connect

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

    2016-09-01

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

  20. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    NASA Astrophysics Data System (ADS)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  1. New instrumental method for determining noble fission gas retained in irradiated nuclear fuels

    SciTech Connect

    Baldwin, D L

    1981-01-01

    The measurement of fission products generated in nuclear fuel is necessary for the complete characterization of the irradiated fuel. The gaseous fission products, xenon and krypton, are of particular importance. A new method has been developed for the measurement of the fission gas retained in nuclear fuel. The method involves extraction of xenon and krypton by melting the fuel in a commercially available furnace. Several factors influence the complete fusion of the fuel and release of the noble gases. Development work aimed at identifying and understanding these factors is discussed. The gases are purified after release from the fuel and collected on cryogenically-cooled activated charcoal. The gases are subsequently released from the charcoal trap and measured by gas chromatography. Column requirements and optimum operating conditions are discussed. Various modifications to the furnace are necessary for reliable performance within the high radiation environment. Other radiological problems are identified and their solutions discussed.

  2. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    SciTech Connect

    Stubbins, James

    2012-12-19

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale.

  3. Fission product behavior during the PBF (Power Burst Facility) Severe Fuel Damage Test 1-1

    SciTech Connect

    Hartwell, J K; Petti, D A; Hagrman, D L; Jensen, S M; Cronenberg, A W

    1987-05-01

    In response to the accident at Three Mile Island Unit 2 (TMI-2), the United States Nuclear Regulatory Commission (USNRC) initiated a series of Severe Fuel Damage tests that were performed in the Power Burst Facility at the Idaho National Engineering Laboratory to obtain data necessary to understand (a) fission product release, transport, and deposition; (b) hydrogen generation; and (c) fuel/cladding material behavior during degraded core accidents. Data are presented about fission product behavior noted during the second experiment of this series, the Severe Fuel Damage Test 1-1, with an in-depth analysis of the fission product release, transport, and deposition phenomena that were observed. Real-time release and transport data of certain fission products were obtained from on-line gamma spectroscopy measurements. Liquid and gas effluent grab samples were collected at selected periods during the test transient. Additional information was obtained from steamline deposition analysis. From these and other data, fission product release rates and total release fractions are estimated and compared with predicted release behavior using current models. Fission product distributions and a mass balance are also summarized, and certain probable chemical forms are predicted for iodine, cesium, and tellurium. An in-depth evaluation of phenomena affecting the behavior of the high-volatility fission products - xenon, krypton, iodine, cesium, and tellurium - is presented. Analysis indicates that volatile release from fuel is strongly influenced by parameters other than fuel temperature. Fission product behavior during transport through the Power Burst Facility effluent line to the fission product monitoring system is assessed. Tellurium release behavior is also examined relatve to the extent of Zircaloy cladding oxidation. 81 fig., 53 tabs.

  4. Thermal stability of fission gas bubble superlattice in irradiated U–10Mo fuel

    SciTech Connect

    Gan, J.; Keiser, D. D.; Miller, B. D.; Robinson, A. B.; Wachs, D. M.; Meyer, M. K.

    2015-09-01

    To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated U-7Mo dispersion and U-10Mo monolithic fuel plates, a FIB-TEM sample of the irradiated U-10Mo fuel with a local fission density of 3.5×1021 fissions/cm3 was used for an in-situ heating TEM experiment. The temperature of the heating holder was raised at a ramp rate of approximately 10 ºC/min up to ~700 ºC, kept at that temperature for about 34 min, continued to 850 ºC with a reduced rate of 5 ºC/min. The result shows a high thermal stability of the fission gas bubble superlattice. The implication of this observation on the fuel microstructural evolution and performance under irradiation is discussed.

  5. Microstructural characterization of irradiated U-7Mo/Al-5Si dispersion fuel to high fission density

    NASA Astrophysics Data System (ADS)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

    2014-11-01

    The fuel development program for research and test reactors calls for improved knowledge on the effect of microstructure on fuel performance in reactors. This paper summarizes the recent TEM microstructural characterization of an irradiated U-7Mo/Al-5Si dispersion fuel plate (R3R050) in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 5.2 × 1021 fissions/cm3. While a large fraction of the fuel grains is decorated with large bubbles, there is no evidence showing interlinking of these bubbles at the specified fission density. The attachment of solid fission product precipitates to the bubbles is likely the result of fission product diffusion into these bubbles. The process of fission gas bubble superlattice collapse appears through bubble coalescence. The results are compared with the previous TEM work on the dispersion fuels irradiated to lower fission density from the same fuel plate.

  6. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    NASA Astrophysics Data System (ADS)

    Pastore, Giovanni; Swiler, L. P.; Hales, J. D.; Novascone, S. R.; Perez, D. M.; Spencer, B. W.; Luzzi, L.; Van Uffelen, P.; Williamson, R. L.

    2015-01-01

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code with a recently implemented physics-based model for fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information in the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior predictions with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, significantly higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  7. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    DOE PAGES

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; ...

    2014-10-12

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertaintymore » in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.« less

  8. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    SciTech Connect

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; Novascone, Stephen R.; Perez, Danielle M.; Spencer, Benjamin W.; Luzzi, Lelio; Uffelen, Paul Van; Williamson, Richard L.

    2014-10-12

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  9. The chemical state of fission products in oxide fuels at different stages of the nuclear fuel cycle

    SciTech Connect

    Kleykamp, H.

    1988-03-01

    A survey of work at the Kernforschungszentrum Karlsruhe is presented on the chemical state of selected fission products that are relevant in the fuel cycle of light water reactor (LWR) and fast breeder reactor fuels. The influence of fuel type and irradiation progress on the composition of the Mo-Tc-Ru-Rh-Pd fission product alloys precipitated in the oxide matrix is examined using the respective multicomponent phase diagrams. The kinetics of dissolution of these phases in nitric acid at the reprocessing stage is discussed. Composition and structure of the residues, and the reprecipitation phenomena from highly active waste (HAW), are elucidated. A second metamorphosis of the fission products is recognized during the vitrification process. The formation of Ru(Rh) oxide and Pd(Rh, U, Te) alloys in simulated vitrified HAW concentrate and in HAW concentrate from the reprocessing of irradiated LWR fuels in interpreted on the basis of heterogeneous equilibria.

  10. Assessment of fission-gas-induced transient swelling in metallic fuel

    SciTech Connect

    Sevy, R H; Cahalan, J E

    1985-03-01

    A model for fission-gas-induced transient swelling in metallic fuel is described. An observation that the strength of metallic fuel becomes very small at a temperature several hundred degrees below the solidus forms the basis for an assumption that, above this temperature, the fuel proceeds through a series of stress-free equilibrium states for a large range of heating rates. Gas bubble coalescence and growth and any effects from ingested sodium are ignored such that the model may tend to underestimate swelling in some circumstances. The fuel swelling model is used to predict the reactivity effect of fission-gas-induced axial expansion of metallic fuel during transient overpower excursions. Comparisons to oxide fuel behavior are made. Sensitivity of results to metallic fuel modeling assumptions are assessed in a parametric study.

  11. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    SciTech Connect

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  12. Fission product release modelling for application of fuel-failure monitoring and detection - An overview

    NASA Astrophysics Data System (ADS)

    Lewis, B. J.; Chan, P. K.; El-Jaby, A.; Iglesias, F. C.; Fitchett, A.

    2017-06-01

    A review of fission product release theory is presented in support of fuel-failure monitoring analysis for the characterization and location of defective fuel. This work is used to describe: (i) the development of the steady-state Visual_DETECT code for coolant activity analysis to characterize failures in the core and the amount of tramp uranium; (ii) a generalization of this model in the STAR code for prediction of the time-dependent release of iodine and noble gas fission products to the coolant during reactor start-up, steady-state, shutdown, and bundle-shifting manoeuvres; (iii) an extension of the model to account for the release of fission products that are delayed-neutron precursors for assessment of fuel-failure location; and (iv) a simplification of the steady-state model to assess the methodology proposed by WANO for a fuel reliability indicator for water-cooled reactors.

  13. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    SciTech Connect

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  14. Mechanistic modelling of urania fuel evolution and fission product migration during irradiation and heating

    NASA Astrophysics Data System (ADS)

    Veshchunov, M. S.; Dubourg, R.; Ozrin, V. D.; Shestak, V. E.; Tarasov, V. I.

    2007-05-01

    The models of the mechanistic code MFPR (Module for Fission Product Release) developed by IBRAE in collaboration with IRSN are described briefly in the first part of the paper. The influence of microscopic defects in the UO2 crystal structure on fission-gas transport out of grains and release from fuel pellets is described. These defects include point defects such as vacancies, interstitials and fission atoms, and extended defects such as bubbles, pores and dislocations. The mechanistic description of chemically active elements behaviour (fission-induced) is based on complex association of diffusion-vaporisation mechanism involving multi-phase and multi-component thermo-chemical equilibrium at the grain boundary with accurate calculation of fuel oxidation. In the second part, results of the code applications are given to different situations: normal LWR reactor operation, high temperature annealing, loss of coolant accident (LOCA) and severe accidents conditions.

  15. Characterization of fission gas bubbles in irradiated U-10Mo fuel

    DOE PAGES

    Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.; ...

    2017-06-06

    A simple, repeatable method for characterization of fission gas bubbles in irradiated U-Mo fuels has been developed. This method involves mechanical potting and polishing of samples along with examination with a scanning electron microscope located outside of a hot cell. The commercially available software packages CellProfiler, MATLAB, and Mathematica are used to segment and analyze the captured images. The results are compared and contrasted. Finally, baseline methods for fission gas bubble characterization are suggested for consideration and further development.

  16. Performance of the fissionTPC and the Potential to Advance the Thorium Fuel Cycle

    NASA Astrophysics Data System (ADS)

    Towell, Rusty; Niffte Collaboration

    2014-09-01

    The NIFFTE fission Time Projection Chamber (fissionTPC) is a powerful tool that is being developed to take precision measurements of neutron-induced fission cross sections of transuranic elements. During the last run at the Los Alamos Neutron Science Center (LANSCE) the fully instrumented TPC took data for the first time. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow cross section systematics to be controlled at the level of 1%. The fissionTPC performance from this run will be shared. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors including the abundance of thorium verses uranium, minimizing radioactive waste, improved reactor safety, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium fueled nuclear reactors will also be discussed.

  17. Correlations for fission product release from N Reactor fuel under high-temperature accident conditions

    SciTech Connect

    Birney, K.R.; Bechtold, D.B.; McCall, T.B.

    1988-03-01

    Empirical correlations were derived for fission product release from metallic uranium alloy 601 N Reactor fuel during postulated accident conditions in which the fuel nears, reaches, or exceeds the melting temperature. The correlations were based on a sparse data base from fuel melted in an inert or steam atmosphere. The empirical correlations are presented for use in subsequent deterministic analyses of N Reactor behavior during hypothetical severe accidents beyond the design basis. 20 refs., 4 figs., 4 tabs.

  18. Validation of ATR Fission Power Deposition Fraction in HEU and LEU Fuel Plates

    SciTech Connect

    G. S. Chang

    2008-09-01

    The Advanced Test Reactor (ATR) is a high power (250 MW), high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of its high power and large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. A detailed plate-by-plate MCNP ATR full core model has been developed and validated for the low-enriched uranium (LEU) fuel conversion feasibility study. Using this model, an analysis has been performed to determine the LEU density and U-235 enrichment required in the fuel meat to yield equivalent K-eff versus effective full power days (EFPDs) between the HEU and LEU cores. This model has also been used to optimize U-235 content of the LEU core, minimizing the differences in K-eff and heat flux profile between the HEU and LEU cores at 115 MW total core power for 125 EFPDs. The LEU core conversion feasibility study evaluated foil type (U-10Mo) fuel with the LEU reference design of 19.7 wt% U-235 enrichment. The LEU reference design has a fixed fuel meat thickness of 0.330 mm and can sustain the same operating cycle length as the HEU fuel. Heat flux and fission power density are parameters that are proportional to the fraction of fission power deposited in fuel. Thus, the accurate determination of the fraction of fission power deposited in the fuel is important to ATR nuclear safety. In this work, a new approach was developed and validated, the Tally Fuel Cells Only (TFCO) method. This method calculates and compares the fission power deposition fraction between HEU and LEU fuel plates. Due to the high density of the U-10Mo LEU fuel, the fission ?-energy deposition fraction is 37.12%, which is larger than the HEU’s ?-energy deposition fraction of 19.7%. As a result, the fuel decay heat cooling will need to be improved

  19. Radiation re-solution of fission gas in non-oxide nuclear fuel

    SciTech Connect

    Matthews, Christopher; Schwen, Daniel; Klein, Andrew C.

    2015-02-01

    Renewed interest in fast nuclear reactors is creating a need for better understanding of fission gas bubble behavior in non-oxide fuels to support very long fuel lifetimes. Collisions between fission fragments and their subsequent cascades can knock fission gas atoms out of bubbles and back into the fuel lattice. We showed that these collisions can be treated as using the so-called ‘‘homogenous’’ atom-by-atom re-solution theory and calculated using the Binary Collision Approximation code 3DOT. The calculations showed that there is a decrease in the re-solution parameter as bubble radius increases until about 50 nm, at which the re-solution parameter stays nearly constant. Furthermore, our model shows ion cascades created in the fuel result in many more implanted fission gas atoms than collisions directly with fission fragments. This calculated re-solution parameter can be used to find a re-solution rate for future bubble simulations.

  20. Fission gas release restrictor for breached fuel rod

    DOEpatents

    Kadambi, N. Prasad; Tilbrook, Roger W.; Spencer, Daniel R.; Schwallie, Ambrose L.

    1986-01-01

    In the event of a breach in the cladding of a rod in an operating liquid metal fast breeder reactor, the rapid release of high-pressure gas from the fission gas plenum may result in a gas blanketing of the breached rod and rods adjacent thereto which impairs the heat transfer to the liquid metal coolant. In order to control the release rate of fission gas in the event of a breached rod, the substantial portion of the conventional fission gas plenum is formed as a gas bottle means which includes a gas pervious means in a small portion thereof. During normal reactor operation, as the fission gas pressure gradually increases, the gas pressure interiorly of and exteriorly of the gas bottle means equalizes. In the event of a breach in the cladding, the gas pervious means in the gas bottle means constitutes a sufficient restriction to the rapid flow of gas therethrough that under maximum design pressure differential conditions, the fission gas flow through the breach will not significantly reduce the heat transfer from the affected rod and adjacent rods to the liquid metal heat transfer fluid flowing therebetween.

  1. Nuclear fuel cycle analysis of the SABR fusion-fission hybrid transmutation reactor

    NASA Astrophysics Data System (ADS)

    Sommer, Chris; Stacey, Weston; Petrovic, Bojan

    2009-11-01

    Various fuel cycles have been designed and analyzed for the Subcritical Advanced Burner Reactor (SABR). SABR is a sodium cooled fast reactor fueled with transuranics (TRU) from spent fuel of light water reactors and driven by a tokamak fusion neutron source based on ITER physics and technology. SABR employs a four batch fuel cycle using an out-to-in shuffling pattern, with the fuel being reprocessed at the end of each cycle. The reprocessing method assumes recovery rates of 99.9% of the actinides and 0.1% of the fission products remain in the recycled fuel. The reprocessing fuel cycles were analyzed to find an optimal cycle length in terms of burn up, power distribution, and materials limitations. Fuel cycles are analyzed using CEA's ERANOS2.0 code, with fuel residence times limited by radiation damage at 100, 150 and 200 dpa.

  2. Fission Product Removal From Spent Oxide Fuel By Head-End Processing

    SciTech Connect

    B. R. Westphal; K. J. Bateman; R. P. Lind; K. L. Howden; G. D. Del Cul

    2005-10-01

    The development of a head-end processing step for spent oxide fuel that applies to both aqueous and pyrometallurgical technologies is being performed by the Idaho National Laboratory, the Oak Ridge National Laboratory, and the Korean Atomic Energy Research Institute through a joint International Nuclear Energy Research Initiative. The processing step employs high temperatures and oxidative gases to promote the oxidation of UO2 to U3O8. Potential benefits of the head-end step include the removal or reduction of fission products as well as separation of the fuel from cladding. Experiments have been performed with irradiated oxide fuel to evaluate the removal of fission products. During these experiments, operating parameters such as temperature and pressure have been varied to discern their effects on the behavior of specific fission products. In general, the extent of removal increases with increasing operating temperature and decreasing pressure. Removal efficiencies as high as 98% have been achieved during testing. Given the results of testing, an explanation of the likely fission product species being removed during the test program is also provided. In addition, experiments have been performed with other oxidative gases (steam and ozone) on surrogates to determine their potential benefit for removal of fission products.

  3. Spent Nuclear Fuel project estimate of volatile fission products release from multi-canister overpacks

    SciTech Connect

    Cooper, T.D.

    1996-08-01

    Spent N-Reactor fuel will be moved from wet pool storage to dry storage at Hanford Washington. This fuel will be sequentially loaded into a Multiple Container Overpack (MCO), moved to the cold vacuum drying station, drained, cold vacuum dried, shipped to the Canister Storage Building (CSB), staged for up to 2 years,hot vacuum dried at 300 degrees C, hot conditioned at 150 degrees C, and finally, sealed and stored for up to 75 years in the CSB.During each proposed process step, the volatile radioactive fission products released to the atmosphere were estimated.Tritium is the only volatile fission product released insignificant amounts during each process step. For an accident scenario involving interior MCO temperature of 600 degrees C for up to 8 hours, it was estimated that many volatile fission products are released.

  4. Transmission electron microscopy characterization of the fission gas bubble superlattice in irradiated U-7 wt%Mo dispersion fuels

    NASA Astrophysics Data System (ADS)

    Miller, B. D.; Gan, J.; Keiser, D. D.; Robinson, A. B.; Jue, J. F.; Madden, J. W.; Medvedev, P. G.

    2015-03-01

    Transmission electron microscopy characterization of irradiated U-7 wt%Mo dispersion fuel were performed on various U-Mo fuel samples to understand the effect of irradiation parameters (fission density, fission rate, and temperature) on the self-organized fission-gas-bubble superlattice that forms in the irradiated U-Mo fuel. The bubble superlattice was seen to form a face centered cubic structure coherent with the host U-7 wt%Mo body-centered cubic structure. At a fission density between 3.0 and 4.5 × 1021 fiss/cm3, the superlattice bubbles appear to have reached a saturation size with additional fission gas associated with increasing burnup predominately accumulating along grain boundaries. At a fission density of ∼4.5 × 1021 fiss/cm3, the U-7 wt%Mo microstructure starts to undergo grain subdivision and can no longer support the ordered bubble superlattice. The sub-divided fuel grains are less than 500 nm in diameter with what appears to be micron-size fission-gas bubbles present on the grain boundaries. Solid fission products typically decorate the inside surface of the micron-sized fission-gas bubbles. Residual superlattice bubbles are seen in areas where fuel grains remain micron sized. Potential mechanisms of the formation and collapse of the bubble superlattice are discussed.

  5. Gaseous fission product management for molten salt reactors and vented fuel systems

    SciTech Connect

    Messenger, S. J.; Forsberg, C.; Massie, M.

    2012-07-01

    Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors (MSRs) and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton (e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years). Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF (78 mole percent) - (HN)F 4 (22 mole percent) with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options

  6. Neutronics of accelerator-driven subcritical fission for burning transuranics in used nuclear fuel

    SciTech Connect

    Sattarov, A.; Assadi, S.; Badgley, K.; Baty, A.; Comeaux, J.; Gerity, J.; Kellams, J.; Mcintyre, P.; Pogue, N.; Sooby, E.; Tsvetkov, P.; Rosaire, G.; Mann, T.

    2013-04-19

    We report the development of a conceptual design for accelerator-driven subcritical fission in a molten salt core (ADSMS). ADSMS is capable of destroying all of the transuranics at the same rate and proportion as they are produced in a conventional nuclear power plant. The ADSMS core is fueled solely by transuranics extracted from used nuclear fuel and reduces its radiotoxicity by a factor 10,000. ADSMS offers a way to close the nuclear fuel cycle so that the full energy potential in the fertile fuels uranium and thorium can be recovered.

  7. Transient Fission Gas Behavior in Uranium Nitride Fuel Under Proposed Space Applications

    DTIC Science & Technology

    1991-12-01

    expansion of the fuel against the cladding , or through excess internal pressure from fission gas which is released to the fuel- clad gap. The former case...Density 96% Cladding NblZr Liner Rhenium Fuel Element Length 46.46 cm Fuel Stack Length 7.62 cm Cladding OD 0.607 cm Cladding Thickness 0.055 cm Liner...operational) the clad surface temperature was approximately 5000K, while at 1 MW (20 MW thermal) the surface temperature jumped to approximately 15000K. 2.3

  8. Grain boundary sweeping and dissolution effects on fission product behavior under severe fuel damage accident conditions

    SciTech Connect

    Rest, J.

    1985-10-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behavior considers the migration and coalescence of fission gas bubbles in either molten uranium, or a zircaloy-uranium eutectic melt. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally irradiated fuel are highlighted.

  9. Measurement and analysis of fission gas release from BNFL's SBR MOX fuel

    NASA Astrophysics Data System (ADS)

    White, R. J.; Fisher, S. B.; Cook, P. M. A.; Stratton, R.; Walker, C. T.; Palmer, I. D.

    2001-01-01

    Puncture results are presented for seven SBR MOX fuel rods from the first prototypical commercial irradiation that was carried out in the Beznau-1 PWR. The rod average burn-up ranged from 31.2 to 35.6 MWd/kgHM. Comparison is made with the percentage of gas released from French MOX fuels and UO 2 fuel. The results show that in the burn-up range investigated, SBR MOX fuel and MIMAS MOX fuel perform similarly, releasing up to about 1% of the fission gas inventory. Comparisons with the Halden Criterion show that SBR MOX has the same release threshold as UO 2 and this suggests that the mechanisms of release in the two fuels are similar. This is further supported by calculations made with the ENIGMA fuel performance code. It is concluded that the apparent differences in fission gas release between SBR MOX and UO 2 fuel, at least in the early stages of release, can be explained by the higher temperatures experienced by MOX fuel.

  10. Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

    DOE PAGES

    Barani, T.; Bruschi, E.; Pizzocri, D.; ...

    2017-01-03

    The modelling of fission gas behaviour is a crucial aspect of nuclear fuel analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. Experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of burst release in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which ismore » applied as an extension of diffusion-based models to allow for the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. Then, the model is assessed for application to integral fuel rod analysis. The approach that we take for model assessment involves implementation in two structurally different fuel performance codes, namely, BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D semi-analytic code). The model is validated against 19 Light Water Reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the qualitative representation of the FGR kinetics and the quantitative predictions of integral fuel rod FGR, relative to the canonical, purely diffusion-based models, with both codes. The overall quantitative improvement of the FGR predictions in the two codes is comparable. Furthermore, calculated radial profiles of xenon concentration are investigated and compared to experimental data, demonstrating the representation of the underlying mechanisms of burst release by the new model.« less

  11. Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

    NASA Astrophysics Data System (ADS)

    Barani, T.; Bruschi, E.; Pizzocri, D.; Pastore, G.; Van Uffelen, P.; Williamson, R. L.; Luzzi, L.

    2017-04-01

    The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release.

  12. Transmission electron microscopy characterization of the fission gas bubble superlattice in irradiated U-7wt% Mo dispersion fuels

    SciTech Connect

    B.D. Miller; J. Gan; D.D. Keiser Jr.; A.B. Robinson; J.-F. Jue; J.W. Madden; P.G. Medvedev

    2015-03-01

    Transmission electron microscopy characterization of irradiated U-7wt% Mo dispersion fuel was performed on various samples to understand the effect of irradiation parameters (fission density, fission rate, and temperature) on the self-organized fission-gas-bubble superlattice that forms in the irradiated U-Mo fuel. The bubble superlattice was seen to form a face-centered cubic structure coherent with the host U-7wt% Mo body centered cubic structure. At a fission density between 3.0 and 4.5 x 1021 fiss/cm3, the superlattice bubbles appear to have reached a saturation size with additional fission gas associated with increasing burnup predominately accumulating along grain boundaries. At a fission density of ~4.5x1021 fiss/cm3, the U-7wt% Mo microstructure undergoes grain subdivision and can no longer support the ordered bubble superlattice. The fuel grains are primarily less than 500 nm in diameter with micron-size fission-gas bubbles present on the grain boundaries. Solid fission products decorate the inside surface of the micron-sized fission-gas bubbles. Residual superlattice bubbles are seen in areas where fuel grains remain micron sized. Potential mechanisms of the formation and collapse of the bubble superlattice are discussed.

  13. A delayed neutron technique for measuring induced fission rates in fresh and burnt LWR fuel

    NASA Astrophysics Data System (ADS)

    Jordan, K. A.; Perret, G.

    2011-04-01

    The LIFE@PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel following re-irradiation in the zero-power PROTEUS research reactor. One such technique utilizes the measurement of delayed neutrons. To demonstrate the feasibility of the delayed neutron technique, fresh and burnt UO 2 fuel samples were irradiated in different positions in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Fission rate ratios of the same sample irradiated in two different positions (inter-positional) and of two different samples irradiated in the same position (inter-sample) were derived from the measurements and compared with Monte Carlo predictions. Derivation of fission rate ratios from the delayed neutron measured signal requires correcting the signal for the delayed neutron source properties, the efficiency of the measurement setup, and the time dependency of the signal. In particular, delayed neutron source properties strongly depend on the fissile and fertile isotopes present in the irradiated sample and must be accounted for when deriving inter-sample fission rate ratios. Measured inter-positional fission rate ratios generally agree within 1σ uncertainty (on the order of 1.0%) with the calculation predictions. For a particular irradiation position, however, a bias of about 2% is observed and is currently under investigation. Calculated and measured inter-sample fission rate ratios have C/E values deviating from unity by less than 1% and within 2σ of the statistical uncertainties. Uncertainty arising from delayed neutron data is also assessed, and is found to give an additional 3% uncertainty factor. The measurement data indicate that uncertainty is overestimated.

  14. Mechanistic approach for nitride fuel evolution and fission product release under irradiation

    NASA Astrophysics Data System (ADS)

    Dolgodvorov, A. P.; Ozrin, V. D.

    2017-01-01

    A model for describing uranium-plutonium mixed nitride fuel pellet burning was developed. Except fission products generating, the model includes impurities of oxygen and carbon. Nitrogen behaviour in nitride fuel was analysed and the nitrogen chemical potential in solid solution with uranium-plutonium nitride was constructed. The chemical program module was tested with the help of thermodynamic equilibrium phase distribution calculation. Results were compared with analogous data in literature, quite good agreement was achieved, especially for uranium sesquinitride, metallic species and some oxides. Calculation of a process of nitride fuel burning was also conducted. Used mechanistic approaches for fission product evolution give the opportunity to find fission gas release fractions and also volumes of intergranular secondary phases. Calculations present that the most massive secondary phases are the oxide and metallic phases. Oxide phase contain approximately 1 % wt of substance over all time of burning with slightly increasing of content. Metallic phase has considerable rising of mass and by the last stage of burning it contains about 0.6 % wt of substance. Intermetallic phase has less increasing rate than metallic phase and include from 0.1 to 0.2 % wt over all time of burning. The highest element fractions of released gaseous fission products correspond to caesium and iodide.

  15. Chemical forms of solid fission products in the irradiated uranium—plutonium mixed nitride fuel

    NASA Astrophysics Data System (ADS)

    Arai, Yasuo; Maeda, Atsushi; Shiozawa, Ken-ichi; Ohmichi, Toshihiko

    1994-06-01

    Chemical forms of solid fission products in the irradiated (U, Pu)N fuel were estimated by both thermodynamic equilibrium calculation and electron microprobe analysis on burnup simulated samples prepared by carbothermic reduction. Besides the MX type matrix phase dissolving zirconium, niobium, yttrium and rare earth elements, the existence of two kinds of inclusion was recognized. One is URu 3 type intermetallic compound constituted by uranium and platinum group elements. The other is an alloy containing molybdenum as a principal constituent. Furthermore, the swelling rate due to solid fission products precipitation was evaluated to be about 0.5% per %FIMA.

  16. Measurement of fission gas release from irradiated U-Mo monolithic fuel samples

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine J.; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 °C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  17. Microstructure of irradiated SBR MOX fuel and its relationship to fission gas release

    NASA Astrophysics Data System (ADS)

    Fisher, S. B.; White, R. J.; Cook, P. M. A.; Bremier, S.; Corcoran, R. C.; Stratton, R.; Walker, C. T.; Ivison, P. K.; Palmer, I. D.

    2002-12-01

    SEM and EPMA examinations of the microstructure and microchemistry of British Nuclear Fuel's quasi-homogeneous SBR MOX fuel following irradiation suggests behaviour which is very similar to that observed in UO 2. Most significantly, a fission gas release of 1% in three-cycle SBR MOX PWR rods is associated with the development of a well-defined intergranular bubble network, which has not been seen previously in the more heterogeneous MOX fuels irradiated under similar conditions. The contrast between the observations is attributed to the relatively low volume fraction and small size of the Pu rich inhomogeneities in the SBR fuel which generate only 4% of the total fission gas and eject most of this into the surrounding mixed oxide matrix. The resulting perturbation in the Xe distribution has a negligible influence on the evolution of the microstructure. A key observation is made from the results of recent post-irradiation annealing experiments performed on SBR MOX and UO 2. These confirm near identical fission gas behaviour in the two fuel types when the influence of thermal conductivity and rod rating are removed.

  18. Measurement of fission gas release from irradiated U–Mo monolithic fuel samples

    SciTech Connect

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine J.; Pool, Karl N.

    2015-06-01

    The uranium–molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium–molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 °C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  19. Measurement of Fission Gas Release from Irradiated U-Mo Monolithic Fuel Samples

    SciTech Connect

    Burkes, Douglas; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of annealing post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1050 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in literature.

  20. Density functional theory calculations of defect and fission gas properties in U-Si fuels

    SciTech Connect

    Andersson, Anders David

    2016-02-03

    Accident tolerant fuels (ATF) are being developed in response to the Fukushima Daiichi accident in Japan. One of the options being pursued is U-Si fuels, such as the U3Si2 and U3Si5 compounds, which benefit from high thermal conductivity (metallic) compared to the UO2 fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for U-Si fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap. The present study investigates point defect and fission gas properties in U3Si2, which is one of the main fuel candidates, using density functional theory (DFT) calculations. Based on a few assumption regarding entropy contributions, defect and fission diffusivities are predicted. Even though uranium silicides have been shown to amorphize easily at low temperature, we assume that U3Si2 remains crystalline under the conditions expected in Light Water Reactors (LWRs). The temperature and dose where amorphization occurs has not yet been well established.

  1. On-site gamma-ray spectroscopic measurements of fission gas release in irradiated nuclear fuel.

    PubMed

    Matsson, I; Grapengiesser, B; Andersson, B

    2007-01-01

    An experimental, non-destructive in-pool, method for measuring fission gas release (FGR) in irradiated nuclear fuel has been developed. Using the method, a significant number of experiments have been performed in-pool at several nuclear power plants of the BWR type. The method utilises the 514 keV gamma-radiation from the gaseous fission product (85)Kr captured in the fuel rod plenum volume. A submergible measuring device (LOKET) consisting of an HPGe-detector and a collimator system was utilised allowing for single rod measurements on virtually all types of BWR fuel. A FGR database covering a wide range of burn-ups (up to average rod burn-up well above 60 MWd/kgU), irradiation history, fuel rod position in cross section and fuel designs has been compiled and used for computer code benchmarking, fuel performance analysis and feedback to reactor operators. Measurements clearly indicate the low FGR in more modern fuel designs in comparison to older fuel types.

  2. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  3. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  4. Behavior of metallic fission products in uranium plutonium mixed oxide fuel

    NASA Astrophysics Data System (ADS)

    Sato, I.; Furuya, H.; Arima, T.; Idemitsu, K.; Yamamoto, K.

    1999-08-01

    Metallic fission products, ruthenium, rhodium, technetium, palladium, and molybdenum, exist in irradiated oxide fuels as metallic inclusions. In this work, the radial distributions of metallic inclusion constituents in the fuel specimen irradiated to a peak burnup of 7-13 at.% were observed with an electron probe microanalysis. Palladium concentration is high at the periphery in all the specimens. Molybdenum shows the same tendency for the 13 at.% burnup specimen. These results showed the significant difference between experimental data and calculations with ORIGEN-2 at such high burnups, which suggested that the migration of palladium and molybdenum was controlled mainly by diffusion of gaseous species containing each metal along the fuel temperature gradient.

  5. Fission Product Release from Molten U/Al Alloy Fuel: A Vapor Transpiration Model

    SciTech Connect

    Whitkop, P.G.

    2001-06-26

    This report describes the application of a vapor transportation model to fission product release data obtained for uranium/aluminum alloy fuel during early Oak Ridge fuel melt experiments. The Oak Ridge data validates the vapor transpiration model and suggests that iodine and cesium are released from the molten fuel surface in elemental form while tellurium and ruthenium are released as oxides. Cesium iodide is postulated to form in the vapor phase outside of the fuel matrix. Kinetic data indicates that cesium iodide can form from Cs atoms and diatomic iodine in the vapor phase. Temperatures lower than those capable of melting fuel are necessary in order to maintain a sufficient I2 concentration. At temperatures near the fuel melting point, cesium can react with iodine atoms to form CsI only on solid surfaces such as aerosols.

  6. Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

    SciTech Connect

    Miao, Yinbin; Mo, Kun; Yacout, Abdellatif; Harp, Jason

    2016-09-15

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

  7. Determination of fission gas release of spent nuclear fuel in puncturing test and in leaching experiments under anoxic conditions

    NASA Astrophysics Data System (ADS)

    González-Robles, E.; Metz, V.; Wegen, D. H.; Herm, M.; Papaioannou, D.; Bohnert, E.; Gretter, R.; Müller, N.; Nasyrow, R.; de Weerd, W.; Wiss, T.; Kienzler, B.

    2016-10-01

    During reactor operation the fission gases Kr and Xe are formed within the UO2 matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO2 fuel itself are widely used as indicators for the release properties of 129I, 137Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H2 overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.

  8. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

  9. Measurement of fission gas release from irradiated UMo dispersion fuel samples

    SciTech Connect

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2016-09-01

    The uranium-molybdenum (U-Mo) alloy dispersed in an Al-Si matrix has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. In this paper, two irradiated samples containing 53.6 vol% U-7wt% Mo fuel particles dispersed in an Al-2wt% Si matrix were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Measurements revealed three distinct fission gas release events for the samples from 400 to 700 oC, as well as a number of minor fission gas releases below and above this temperature range. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature with exceptional agreement.

  10. Measurement of fission gas release from irradiated Usbnd Mo dispersion fuel samples

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2016-09-01

    The uranium-molybdenum (Usbnd Mo) alloy dispersed in an Alsbnd Si matrix has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. In this paper, two irradiated samples containing 53.9 vol% U-7wt% Mo fuel particles dispersed in an Al-2wt% Si matrix were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Measurements revealed three distinct fission gas release events for the samples from 400 to 700 °C, as well as a number of minor fission gas releases below and above this temperature range. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature with exceptional agreement.

  11. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    SciTech Connect

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

  12. Using New Fission Data with the Multi-detector Analysis System for Spent Nuclear Fuel

    SciTech Connect

    A. V. Ramayya; A.V. Daniel; C. J. Beyer; E. L. Reber; G. M. Ter-Akopian; G.S. Popeko; J. D. Cole; J. H. Hamilton; J. K. Jewell; M. W. Drigert; R. Aryaeinejad; Ts.Yu. Oganessian

    1998-11-01

    New experiments using an array of high purity germanium detectors and fast liquid scintillation detectors has been performed to observe the radiation emitted from the induced fission of 235U with a beam of thermal neutrons. The experiment was performed at the Argonne National Laboratory Intense Pulsed Neutron Source. Preliminary observations of the data are presented. A nondestructive analysis system for the characterization of DOE spent nuclear fuel based on these new data is presented.

  13. Using New Fission Data with the Multi-detector Analysis System for Spent Nuclear Fuel

    SciTech Connect

    Cole, Jerald Donald

    1998-11-01

    New experiments using an array of high purity germanium detectors and fast liquid scintillation detectors has been performed to observe the radiation emitted from the induced fission of 235U with a beam of thermal neutrons. The experiment was performed at the Argonne National Laboratory Intense Pulsed Neutron Source. Preliminary observations of the data are presented. A nondestructive analysis system for the characterization of DOE spent nuclear fuel based on these new data is presented.

  14. A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

    SciTech Connect

    Bartram, B.W.; Dougherty, D.K.

    1987-01-01

    This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs. (TEM)

  15. Separation of the rare-earth fission product poisons from spent nuclear fuel

    SciTech Connect

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

  16. Determination of actinide and fission-product isotopes in very-high-burnup spent nuclear fuel.

    SciTech Connect

    Sullivan, V. S.; Bowers, D. L.; Clark, M. A.; Graczyk, D. G.; Tsai, Y.; Streets, W. E.; Vander Pol, M. H.; Billone, M. C.

    2008-07-01

    A work plan was desired that would produce data for a wide array of actinide and fission-product isotopes with reasonably good accuracy and relatively low cost. An analysis scheme involving a fairly small number of separations, dilutions, and measurement methods was used to generate information on 74 isotopes in two spent-fuel samples of >70 GWd/MTU burnup. Some of the measured isotopes are of high interest for burnup-credit evaluations and had not been reported previously for high-burnup fuels.

  17. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    SciTech Connect

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W; Hagrman, D L; McClure, P R; Leonard, M T

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  18. Characterization of intergranular fission gas bubbles in U-Mo fuel.

    SciTech Connect

    Kim, Y. S.; Hofman, G.; Rest, J.; Shevlyakov, G. V.; Nuclear Engineering Division; SSCR RIAR

    2008-04-14

    This report can be divided into two parts: the first part, which is composed of sections 1, 2, and 3, is devoted to report the analyses of fission gas bubbles; the second part, which is in section 4, is allocated to describe the mechanistic model development. Swelling data of irradiated U-Mo alloy typically show that the kinetics of fission gas bubbles is composed of two different rates: lower initially and higher later. The transition corresponds to a burnup of {approx}0 at% U-235 (LEU) or a fission density of {approx}3 x 10{sup 21} fissions/cm{sup 3}. Scanning electron microscopy (SEM) shows that gas bubbles appear only on the grain boundaries in the pretransition regime. At intermediate burnup where the transition begins, gas bubbles are observed to spread into the intragranular regions. At high burnup, they are uniformly distributed throughout fuel. In highly irradiated U-Mo alloy fuel large-scale gas bubbles form on some fuel particle peripheries. In some cases, these bubbles appear to be interconnected and occupy the interface region between fuel and the aluminum matrix for dispersion fuel, and fuel and cladding for monolithic fuel, respectively. This is a potential performance limit for U-Mo alloy fuel. Microscopic characterization of the evolution of fission gas bubbles is necessary to understand the underlying phenomena of the macroscopic behavior of fission gas swelling that can lead to a counter measure to potential performance limit. The microscopic characterization data, particularly in the pre-transition regime, can also be used in developing a mechanistic model that predicts fission gas bubble behavior as a function of burnup and helps identify critical physical properties for the future tests. Analyses of grain and grain boundary morphology were performed. Optical micrographs and scanning electron micrographs of irradiated fuel from RERTR-1, 2, 3 and 5 tests were used. Micrographic comparisons between as-fabricated and as-irradiated fuel revealed

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

    SciTech Connect

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

    2013-09-01

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

  20. Fuel age impacts on gaseous fission product capture during separations

    SciTech Connect

    Jubin, Robert T.; Soelberg, Nicolas R.; Strachan, Denis M.; Ilas, G.

    2012-09-21

    As a result of fuel reprocessing, volatile radionuclides will be released from the facility stack if no processes are put in place to remove them. The radionuclides that are of concern in this document are 3H, 14C, 85Kr, and 129 Rosnick 2007 I. The question we attempt to answer is how efficient must this removal process be for each of these radionuclides? To answer this question, we examine the three regulations that may impact the degree to which these radionuclides must be reduced before process gases can be released from the facility. These regulations are 40 CFR 61 (EPA 2010a), 40 CFR 190(EPA 2010b), and 10 CFR 20 (NRC 2012), and they apply to the total radonuclide release and to the dose to a particular organ – the thyroid. Because these doses can be divided amongst all the radionuclides in different ways and even within the four radionuclides in question, several cases are studied. These cases consider for the four analyzed radionuclides inventories produced for three fuel types—pressurized water reactor uranium oxide (PWR UOX), pressurized water reactor mixed oxide (PWR MOX), and advanced high-temperature gascooled reactor (AHTGR)—several burnup values and time out of reactor extending to 200 y. Doses to the maximum exposed individual (MEI) are calculated with the EPA code CAP-88 ( , 1992). Two dose cases are considered. The first case, perhaps unrealistic, assumes that all of the allowable dose is assigned to the volatile radionuclides. In lieu of this, for the second case a value of 10% of the allowable dose is arbitrarily selected to be assigned to the volatile radionuclides. The required decontamination factors (DFs) are calculated for both of these cases, including the case for the thyroid dose for which 14C and 129I are the main contributors. However, for completeness, for one fuel type and burnup, additional cases are provided, allowing 25% and 50% of the allowable dose to be assigned to the volatile radionuclides. Because 3H and 85Kr have

  1. Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel

    SciTech Connect

    Lillo, T. M.; Rooyen, I. J.

    2016-02-26

    The relationship between grain boundary character and fission product migration is identified as an important knowledge gap in order to advance the understanding of fission product release from TRISO fuel particles. Precession electron diffraction (PED), a TEM-based technique, was used in this study to quickly and efficiently provide the crystallographic information needed to identify grain boundary misorientation, grain boundary type (low or high angle) and whether the boundary is coincident site lattice (CSL) – related, in irradiated SiC. Analysis of PED data showed the grain structure of the SiC layer in an irradiated TRISO fuel particle from the AGR-1 experiment to be composed mainly of twin boundaries with a small fraction of low angle grain boundaries (<10%). In general, fission products favor precipitation on random, high angle grain boundaries but can precipitate out on low angle and CSL-related grain boundaries to a limited degree. Pd is capable of precipitating out on all types of grain boundaries but most prominently on random, high angle grain boundaries. Pd-U and Pd-Ag precipitates were found on CSL-related as well as random high angle grain boundaries but not on low angle grain boundaries. In contrast, precipitates containing only Ag were found only on random, high angle grain boundaries but not on either low angle or CSL-related grain boundaries.

  2. Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel

    DOE PAGES

    Lillo, T. M.; Rooyen, I. J.

    2016-02-26

    The relationship between grain boundary character and fission product migration is identified as an important knowledge gap in order to advance the understanding of fission product release from TRISO fuel particles. Precession electron diffraction (PED), a TEM-based technique, was used in this study to quickly and efficiently provide the crystallographic information needed to identify grain boundary misorientation, grain boundary type (low or high angle) and whether the boundary is coincident site lattice (CSL) – related, in irradiated SiC. Analysis of PED data showed the grain structure of the SiC layer in an irradiated TRISO fuel particle from the AGR-1 experimentmore » to be composed mainly of twin boundaries with a small fraction of low angle grain boundaries (<10%). In general, fission products favor precipitation on random, high angle grain boundaries but can precipitate out on low angle and CSL-related grain boundaries to a limited degree. Pd is capable of precipitating out on all types of grain boundaries but most prominently on random, high angle grain boundaries. Pd-U and Pd-Ag precipitates were found on CSL-related as well as random high angle grain boundaries but not on low angle grain boundaries. In contrast, precipitates containing only Ag were found only on random, high angle grain boundaries but not on either low angle or CSL-related grain boundaries.« less

  3. Radioactive Fission Product Release from Defective Light Water Reactor Fuel Elements

    SciTech Connect

    Konyashov, Vadim V.; Krasnov, Alexander M.

    2002-04-15

    Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. An approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)

  4. Partitioning of selected fission products from irradiated oxide fuel induced by thermal treatment

    NASA Astrophysics Data System (ADS)

    Shcherbina, Natalia; Kivel, Niko; Günther-Leopold, Ines

    2013-06-01

    The release of fission products (FPs) from spent nuclear fuel (SNF) has been studied as a function of the temperature and redox conditions. The present paper concerns essentially the high temperature separation of Cs and Sr from irradiated pressurized (PWR) and boiling water reactor (BWR) fuel of different burn-up levels with use of an in-house designed system for inductive vaporization (InVap). Using thermodynamic calculations with the Module of Fission Product Release (MFPR) code along with annealing experiments on SNF in the InVap it was shown that the speciation of Cs and Sr, hence their release behavior at high temperature, is sensitive to the redox conditions during thermal treatment. It was demonstrated that annealing conditions in the InVap can be adjusted in the way to promote the release of selected FPs without significant loss of the fuel matrix or actinides: complete release of Cs and I was achieved during treatment of irradiated fuel at 1800 °C under reducing atmosphere (0.7% H2/Ar mixture). The developed partitioning procedure can be used for the SNF pretreatment as an advanced head-end step in the hydrometallurgical or pyrochemical reprocessing technology.

  5. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    SciTech Connect

    Jason M. Harp; Paul A. Demkowicz

    2014-10-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10-4 to 10-5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.

  6. Post irradiation examination of simulated fission product doped hyperstoichiometric mixed oxide fuel pins*1

    NASA Astrophysics Data System (ADS)

    Götzmann, O.; Kleykamp, H.

    1980-03-01

    Two miniature fuel pins containing uranium-plutonium oxide with a hyperstoichiometric oxygen-to-metal ratio and selective fission product elements have been irradiated in the BR 2 reactor at Mol, Belgium, for two reactor cycles (46 days). One of the pins had a niobium metal coating on the inner cladding surface to act as oxygen getter. Both pins were subjected to a detailed examination by ceramography and electronprobe microanalysis. The results have been interpreted in the light of a recently published thermochemical model for the cladding attack. The very different oxygen potential environments in the two pins produced entirely different clad corrosion phenomena probably due to different cladding attack mechanisms. The niobium coating worked well in reducing the oxygen potential. However, there exists a draw back with niobium due to the formation of relatively stable intermetallic phases with noble metal fission products.

  7. Accelerator-driven subcritical fission in molten salt core: Closing the nuclear fuel cycle for green nuclear energy

    SciTech Connect

    McIntyre, Peter; Assadi, Saeed; Badgley, Karie; Baker, William; Comeaux, Justin; Gerity, James; Kellams, Joshua; McInturff, Al; Pogue, Nathaniel; Sattarov, Akhdiyor; Sooby, Elizabeth; Tsvetkov, Pavel; Phongikaroon, Supathorn; Simpson, Michael

    2013-04-19

    A technology for accelerator-driven subcritical fission in a molten salt core (ADSMS) is being developed as a basis for the destruction of the transuranics in used nuclear fuel. The molten salt fuel is a eutectic mixture of NaCl and the chlorides of the transuranics and fission products. The core is driven by proton beams from a strong-focusing cyclotron stack. This approach uniquely provides an intrinsically safe means to drive a core fueled only with transuranics, thereby eliminating competing breeding terms.

  8. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, 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.

  9. Method and device for fabricating dispersion fuel comprising fission product collection spaces

    DOEpatents

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

    A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

  10. Fission product partitioning in aerosol release from simulated spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Di Lemma, F. G.; Colle, J. Y.; Rasmussen, G.; Konings, R. J. M.

    2015-10-01

    Aerosols created by the vaporization of simulated spent nuclear fuel (simfuel) were produced by laser heating techniques and characterised by a wide range of post-analyses. In particular attention has been focused on determining the fission product behaviour in the aerosols, in order to improve the evaluation of the source term and consequently the risk associated with release from spent fuel sabotage or accidents. Different simulated spent fuels were tested with burn-up up to 8 at. %. The results from the aerosol characterisation were compared with studies of the vaporization process by Knudsen Effusion Mass Spectrometry and thermochemical equilibrium calculations. These studies permit an understanding of the aerosol gaseous precursors and the gaseous reactions taking place during the aerosol formation process.

  11. MICRO/NANO-STRUCTURAL EXAMINATION AND FISSION PRODUCT IDENTIFICATION IN NEUTRON IRRADIATED AGR-1 TRISO FUEL

    SciTech Connect

    van Rooyen, I. J.; Lillo, T. M.; Wen, H. M.; Hill, C. M.; Holesinger, T. G.; Wu, Y. Q.; Aguiara, J. A.

    2016-11-01

    Advanced microscopic and microanalysis techniques were developed and applied to study irradiation effects and fission product behavior in selected low-enriched uranium oxide/uranium carbide TRISO-coated particles from fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA. Although no TRISO coating failures were detected during the irradiation, the fraction of Ag-110m retained in individual particles often varied considerably within a single compact and at the capsule level. At the capsule level Ag-110m release fractions ranged from 1.2 to 38% and within a single compact, silver release from individual particles often spanned a range that extended from 100% retention to nearly 100% release. In this paper, selected irradiated particles from Baseline, Variant 1 and Variant 3 type fueled TRISO coated particles were examined using Scanning Electron Microscopy, Atom Probe Tomography; Electron Energy Loss Spectroscopy; Precession Electron Diffraction, Transmission Electron Microscopy, Scanning Transmission Electron Microscopy (STEM), High Resolution Electron Microscopy (HRTEM) examinations and Electron Probe Micro-Analyzer. Particle selection in this study allowed for comparison of the fission product distribution with Ag retention, fuel type and irradiation level. Nano sized Ag-containing features were predominantly identified in SiC grain boundaries and/or triple points in contrast with only two sitings of Ag inside a SiC grain in two different compacts (Baseline and Variant 3 fueled compacts). STEM and HRTEM analysis showed evidence of Ag and Pd co-existence in some cases and it was found that fission product precipitates can consist of multiple or single phases. STEM analysis also showed differences in precipitate compositions between Baseline and Variant 3 fuels. A higher density of fission product precipitate clusters were identified in the SiC layer in particles from the Variant 3 compact compared with the Variant 1 compact. Trend analysis shows

  12. Analyzing Losses: Transuranics into Waste and Fission Products into Recycled Fuel

    SciTech Connect

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

    2010-11-01

    All mass streams from separations and fuel fabrication are products that must meet criteria. Those headed for disposal must meet waste acceptance criteria (WAC) for the eventual disposal sites corresponding to their waste classification. Those headed for reuse must meet fuel or target impurity limits. A “loss” is any material that ends up where it is undesired. The various types of losses are linked in the sense that as the loss of transuranic (TRU) material into waste is reduced, often the loss or carryover of waste into TRU or uranium is increased. We have analyzed four separation options and two fuel fabrication options in a generic fuel cycle. The separation options are aqueous uranium extraction plus (UREX+1), electrochemical, Atomics International reduction oxidation separation (AIROX), and melt refining. UREX+1 and electrochemical are traditional, full separation techniques. AIROX and melt refining are taken as examples of limited separations, also known as minimum fuel treatment. The fuels are oxide and metal. To define a generic fuel cycle, a fuel recycling loop is fed from used light water reactor (LWR) uranium oxide fuel (UOX) at 51 MWth-day/kg-iHM burnup. The recycling loop uses a fast reactor with TRU conversion ratio (CR) of 0.50. Excess recovered uranium is put into storage. Only waste, not used fuel, is disposed – unless the impurities accumulate to a level so that it is impossible to make new fuel for the fast reactor. Impurities accumulate as dictated by separation removal and fission product generation. Our model approximates adjustment to fast reactor fuel stream blending of TRU and U products from incoming LWR UOX and recycling FR fuel to compensate for impurity accumulation by adjusting TRU:U ratios. Our mass flow model ignores postulated fuel impurity limits; we compare the calculated impurity values with those limits to identify elements of concern. AIROX and melt refining cannot be used to separate used LWR UOX-51 because they cannot

  13. A physical description of fission product behavior fuels for advanced power reactors.

    SciTech Connect

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  14. Chemical thermodynamics of Cs and Te fission product interactions in irradiated LMFBR mixed-oxide fuel pins

    NASA Astrophysics Data System (ADS)

    Adamson, M. G.; Aitken, E. A.; Lindemer, T. B.

    1985-02-01

    A combination of fuel chemistry modelling and equilibrium thermodynamic calculations has been used to predict the atom ratios of Cs and Te fission products (Cs:Te) that find their way into the fuel-cladding interface region of irradiated stainless steel-clad mixed-oxide fast breeder reactor fuel pins. It has been concluded that the ratio of condensed, chemically-associated Cs and Te in the interface region,Čs:Te, which in turn determines the Te activity, is controlled by an equilibrium reaction between Cs 2Te and the oxide fuel, and that the value of Čs:Te is, depending on fuel 0:M, either equal to or slightly less than 2:1. Since Cs and Te fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), the observed out-of-pile Cs:Te thresholds for FCCI (4˜:1) and FPLME (2˜:1) have been rationalized in terms of Cs:Te thermochemistry and phase equilibria. Also described in the paper is an updated chemical evolution model for reactive/volatile fission product behavior in irradiated oxide pins.

  15. Effect of fission rate on the microstructure of coated UMo dispersion fuel

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Parthoens, Y.; Cornelis, G.; Kuzminov, V.; Koonen, E.; Van den Berghe, S.; Ye, B.; Hofman, G. L.; Schulthess, Jason

    2017-10-01

    Compared to previous irradiation experiments containing UMo/Al dispersion fuel plates, the SELENIUM irradiation experiment performed at the SCK·CEN BR2 reactor in 2012 showed an improved plate swelling behavior. However, in the high burn-up area of the plates a significant increase in meat thickness was still measured. The origin of this increase is currently not firmly established, but it is clear from the observed microstructure that the swelling rate still is too high for practical purposes and needs to be reduced. It was stipulated that the swelling occurred at the high burnup areas which are also the high power zones at beginning of life. For that reason, an experiment was proposed to investigate the influence of fission rate (i.e. power) on some of the observed phenomena. For this purpose, a sibling plate to a high power (BOL>470 W/cm2) SELENIUM plate was irradiated during four BR2 cycles. The SELENIUM 1a fuel plate was submitted to a local maximum heat flux below 350 W/cm2, throughout the full irradiation. At the end of the last cycle, the SELENIUM 1a fuel plate reached a maximum local burnup value of close to 75%235U compared to 70%235U for the SELENIUM high power plates. When comparing to the results on the SELENIUM plates, the non-destructive tests clearly show a continued linear swelling behavior of the low power irradiated fuel plate SELENIUM 1a in the high burn-up region. The influence of the fission rate is also evidenced in the microstructural examination of the fuel showing that there is no formation of interaction layer at the high burn-up region.

  16. Fission product release and survivability of UN-kernel LWR TRISO fuel

    SciTech Connect

    T. M. Besmann; M. K. Ferber; H.-T. Lin; B. P. Collin

    2014-05-01

    A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from fission product recoil calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 um diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated within a TRISO particle undergoing burnup. Creep and swelling of the inner and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by computing the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers from internal pressure and thermomechanics of the layers. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.

  17. On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel

    SciTech Connect

    Perret, G.; Jordan, K. A.

    2011-07-01

    Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)

  18. State-of-the-art on instant release of fission products from spent nuclear fuel

    SciTech Connect

    Kienzler, Bernhard; Gonzalez-Robles, Ernesto

    2013-07-01

    Within the EURATOM FP7 Collaborative Project 'Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)', a State-of-the-Art Report was prepared. The fast / instant release fraction (IRF) is defined as a fraction of the inventory of radionuclides that may be rapidly released from the fuel and fuel assembly materials at the time of canister breaching. In the context of safety analysis for a repository, the time span for mobilization of this fraction can be considered instantaneously, even if the process takes some time in experiments. Radionuclides contributing to the fast release are fission gases (Kr and Xe), easily soluble elements such as cesium and iodine, and other elements which are hardly incorporated in the UO{sub 2} crystal lattice. The present contribution summarizes the results obtained from published studies focused on rapid release experiments carried out with different spent nuclear fuel (SNF), samples, sizes, techniques (batch and flow-through), and durations. A total of 80 experiments cover the study of UO{sub 2} SNF from pressure water reactors (PWR) of different initial enrichments and burn-up, while 20 experiments were performed with UO{sub 2} SNF from boiling water reactors (BWR) and 8 with MOX fuel. (authors)

  19. Partitioning of fission products from irradiated nitride fuel using inductive vaporization

    SciTech Connect

    Shcherbina, N.; Kulik, D.A.; Kivel, N.; Potthast, H.D.; Guenther-Leopold, I.

    2013-07-01

    Irradiated nitride fuel (Pu{sub 0.3}Zr{sub 0.7})N fabricated at PSI in frame of the CONFIRM project and having a burn-up of 10.4 % FIMA (Fission per Initial Metal Atom) has been investigated by means of inductive vaporization. The study of thermal stability and release behavior of Pu, Am, Zr and fission products (FPs) was performed in a wide temperature range (up to 2300 C. degrees) and on different redox conditions. On-line monitoring by ICP-MS detected low nitride stability and significant loss of Pu and Am at T>1900 C. degrees during annealing under inert atmosphere (Ar). The oxidative pre-treatment of nitride fuel on air at 1000 C. degrees resulted in strong retention of Pu and Am in the solid, as well as of most FPs. Thermodynamic modelling of elemental speciation using GEM-Selektor v.3 code (Gibbs Energy Minimization Selektor), supported by a comprehensive literature review on thermodynamics of actinides and FPs, revealed a number of binary compounds of Cs, Mo, Te, Sr and Ba to occur in the solid. Speciation of some FPs in the fuel is discussed and compared to earlier results of electron probe microanalysis (EPMA). Predominant vapor species predicted by GEM-Selektor calculations were Pu(g), Am(g) and N{sub 2}. Nitrogen can be completely released from the fuel after complete oxidation at 1000 C. degrees. With regard to the irradiated nitride reprocessing technology, this result can have an important practical application as an alternative way for {sup 15}N recovery. (authors)

  20. The use of gamma spectrometry in the determination of fission products migration in irradiated fuel

    SciTech Connect

    2015-07-01

    Non destructive examinations realized in hot cells of LECA STAR facility give main data on irradiated fuel rods and pins. Among those examinations, gamma spectrometry allows access to fuel, inside the cladding, thanks to gamma rays of fission products such as {sup 137}Cs, {sup 154}Eu coming from pellets... From those gamma scannings we can detect the position and the length of the fuel column and of its pellets, calculate local burnup.. In the database of our lab we have already such gamma scannings on hundreds of rods or pins with different fuels, claddings and irradiations conditions (under nominal or non nominal). We have detected that in specific cases, an unusual shape of the {sup 137}Cs scanning (distribution quite different from those of {sup 154}Eu) can be explained by the migration of this isotope, moving to the cold sides of the pellet. This phenomenon is mainly associated with an increasing of the pellet's temperature. Based on our observations, we have developed a quantitative approach of the changes on the {sup 137}Cs scannings through the calculation of appropriate 'indicators'. Those calculations allow us to be able to localize, quantify and compare the {sup 137}Cs migration all along rods or pins. Those migration results are so quickly and easily achievable from gamma measurements and can then be easily correlated to other observations realized with destructive examinations, puncturing and calculations, already realized or to come. Because such a migration is the result of temperature increasing in the pellets, our indicators can be directly associated with this local temperature. In perspective, thermically activated phenomena such as geometrical changes in the shape of the pellets, fission gas release... can also indirectly be deduced from our indicators. (authors)

  1. Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

    NASA Astrophysics Data System (ADS)

    Acır, Adem; Altunok, Taner

    2010-10-01

    HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li20Sn80 coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.

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

    SciTech Connect

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

    1993-12-31

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

  3. Neutronics Design of a Thorium-Fueled Fission Blanket for LIFE (Laser Inertial Fusion-based Energy)

    SciTech Connect

    Powers, J; Abbott, R; Fratoni, M; Kramer, K; Latkowski, J; Seifried, J; Taylor, J

    2010-03-08

    The Laser Inertial Fusion-based Energy (LIFE) project at LLNL includes development of hybrid fusion-fission systems for energy generation. These hybrid LIFE engines use high-energy neutrons from laser-based inertial confinement fusion to drive a subcritical blanket of fission fuel that surrounds the fusion chamber. The fission blanket contains TRISO fuel particles packed into pebbles in a flowing bed geometry cooled by a molten salt (flibe). LIFE engines using a thorium fuel cycle provide potential improvements in overall fuel cycle performance and resource utilization compared to using depleted uranium (DU) and may minimize waste repository and proliferation concerns. A preliminary engine design with an initial loading of 40 metric tons of thorium can maintain a power level of 2000 MW{sub th} for about 55 years, at which point the fuel reaches an average burnup level of about 75% FIMA. Acceptable performance was achieved without using any zero-flux environment 'cooling periods' to allow {sup 233}Pa to decay to {sup 233}U; thorium undergoes constant irradiation in this LIFE engine design to minimize proliferation risks and fuel inventory. Vast reductions in end-of-life (EOL) transuranic (TRU) inventories compared to those produced by a similar uranium system suggest reduced proliferation risks. Decay heat generation in discharge fuel appears lower for a thorium LIFE engine than a DU engine but differences in radioactive ingestion hazard are less conclusive. Future efforts on development of thorium-fueled LIFE fission blankets engine development will include design optimization, fuel performance analysis work, and further waste disposal and nonproliferation analyses.

  4. TEM Characterization of U-7Mo/Al-2Si Dispersion Fuel Irradiated to Intermediate and High Fission Densities

    SciTech Connect

    J. Gan; D.D. Keiser, Jr.; B.D. Miller; A.B. Robinson; J-F. Jue; P.G. Medvedev; D.M. Wachs

    2012-05-01

    This paper will discuss the results of TEM analysis that was performed on two samples taken from the low flux and high flux sides of the fuel plate with U-7Mo fuel particles dispersed in U-2Si matrix. The corresponding local fission density of the fuel particles and the peak fuel plate centerline temperature between the low flux and high flux samples are 3.32 x 10{sup 27} f/m{sup 3} and 90 C, and 6.31 x 10{sup 27} f/m{sup 3} and 120 C, respectively. The results of this work showed the presence of a bubble superlattice within the U-7Mo grains that accommodated fission gases (e.g., Xe). The presence of this structure helps the U-7Mo exhibit a stable swelling behavior during irradiation. The Si-rich interaction layers that develop around the fuel particles at the U-7Mo/matrix interface during fuel plate fabrication and irradiation become amorphous during irradiation. The change in bubble distribution at the high fission density suggests that the bubble superlattice is stable as the U-7Mo matrix remains crystalline. It appears that there is a threshold Si content in the fuel particle above which the U-Mo turns to amorphous under irradiation. The threshold Si content is approximately 8 at.% and 4 at.% for low flux and high flux condition, respectively.

  5. Application of U10Mo Fuel for Space Fission Power Applications - White Paper

    SciTech Connect

    James Werner

    2014-07-01

    A novel reactor design has been proposed for space applications to provide hundreds of watts to one or two kilowatts of electrical power. The reactor concept proposed uses the alloy U10Mo (uranium with 10 weight percent molybdenum) as the fuel. This fuel was selected for its high uranium density, high thermal conductivity, and excellent neutronic characteristics for this application. The core is surrounded by a BeO reflector. Heat is carried from the reactor by liquid metal heat pipes. A shadow shield of LiH tungsten is also utilized to reduce the neutron and gamma radiation dose to the rest of the spacecraft. This design represents a best effort at minimizing the complexity of the fission system and reducing the mass of the system. The compact nature of the block UMo core and BeO radial reflector allows the reactor diameter to be as small as practical while still meeting the neutronic and thermal power demands. This directly results in a reduced shield mass since the reactor diameter dictates the footprint of the radiation shield. The use of heat pipes offers a straightforward primary heat transport approach using proven liquid-metal heat pipe technology. Further, the elimination of a liquid core coolant system heat transport components, both at the reactor side and radiator side, contributes to reducing the total part-count and lowering system mass. The proposed reactor is using a fuel that is being developed by DOE, but there are significant differences in the fuels enrichment, operating conditions and the physical shape of the fuel itself. This paper attempts to highlight some of the basic consideration and needs that would be expected to be met in developing this fuel and qualifying it for use.

  6. Molten Salt Fuel Version of Laser Inertial Fusion Fission Energy (LIFE)

    SciTech Connect

    Moir, R W; Shaw, H F; Caro, A; Kaufman, L; Latkowski, J F; Powers, J; Turchi, P A

    2008-10-24

    Molten salt with dissolved uranium is being considered for the Laser Inertial Confinement Fusion Fission Energy (LIFE) fission blanket as a backup in case a solid-fuel version cannot meet the performance objectives, for example because of radiation damage of the solid materials. Molten salt is not damaged by radiation and therefore could likely achieve the desired high burnup (>99%) of heavy atoms of {sup 238}U. A perceived disadvantage is the possibility that the circulating molten salt could lend itself to misuse (proliferation) by making separation of fissile material easier than for the solid-fuel case. The molten salt composition being considered is the eutectic mixture of 73 mol% LiF and 27 mol% UF{sub 4}, whose melting point is 490 C. The use of {sup 232}Th as a fuel is also being studied. ({sup 232}Th does not produce Pu under neutron irradiation.) The temperature of the molten salt would be {approx}550 C at the inlet (60 C above the solidus temperature) and {approx}650 C at the outlet. Mixtures of U and Th are being considered. To minimize corrosion of structural materials, the molten salt would also contain a small amount ({approx}1 mol%) of UF{sub 3}. The same beryllium neutron multiplier could be used as in the solid fuel case; alternatively, a liquid lithium or liquid lead multiplier could be used. Insuring that the solubility of Pu{sup 3+} in the melt is not exceeded is a design criterion. To mitigate corrosion of the steel, a refractory coating such as tungsten similar to the first wall facing the fusion source is suggested in the high-neutron-flux regions; and in low-neutron-flux regions, including the piping and heat exchangers, a nickel alloy, Hastelloy, would be used. These material choices parallel those made for the Molten Salt Reactor Experiment (MSRE) at ORNL. The nuclear performance is better than the solid fuel case. At the beginning of life, the tritium breeding ratio is unity and the plutonium plus {sup 233}U production rate is {approx}0

  7. Fission Product Release and Survivability of UN-Kernel LWR TRISO Fuel

    SciTech Connect

    Besmann, Theodore M; Ferber, Mattison K; Lin, Hua-Tay

    2014-01-01

    A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from range calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 m diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated with a TRISO particle as a function of fluence. Creep and swelling of the inner and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by measuring the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers as a function of fluence. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.

  8. Molecular Dynamics Simulation of Fission Fragment Damage in Nuclear Fuel and Surrogate Material

    SciTech Connect

    Devanathan, Ram

    2017-01-01

    ABSTRACT

    We have performed classical molecular dynamics simulations of swift heavy ion damage, typical of fission fragments, in nuclear fuel (UO2) for energy deposition per unit length of 3.9 keV/nm. We did not observe amorphization. The damage mainly consisted of isolated point defects. Only about 1% of the displacements occur on the uranium sublattice. Oxygen Frenkel pairs are an order of magnitude more numerous than uranium Frenkel pairs in the primary damage state. In contrast, previous results show that the ratio of Frenkel pairs on the two sublattices is close to the stoichiometric ratio in ceria. These differences in the primary damage state may lead to differences in radiation response of UO2and CeO2.

  9. Simulated fission product-SiC interaction in Triso-coated LEU or MEU HTGR fuel particles

    SciTech Connect

    Pearson, R.L.; Lindemer, T.B.; Beahm, E.C.

    1980-11-01

    Proliferation issues relating to the use of highly enriched uranium (HEU) have led to an evaluation of the fission product-SiC interaction problems that might arise if low enriched uranium (LEU) or medium enriched uranium (MEU) were used as fissile fuel in HTGR systems. Simulated Triso-coated UO/sub 2/, UC/sub 2/, and UO/sub 2//UC/sub 2/ particles mixed with varying amounts of Mo, Ru, Rh, Pd, Ag, and Cd were prepared. These fission products were chosen because, after full burnup, their concentrations are higher in LEU and MEU fuels than in HEU fuel. After the particles were heat treated in the laboratory, their behavior was examined by use of metallography, scanning electron microscopy, and electron microprobe x-ray analysis.

  10. A model for the influence of microstructure, precipitate pinning and fission gas behavior on irradiation-induced recrystallization of nuclear fuels

    NASA Astrophysics Data System (ADS)

    Rest, J.

    2004-03-01

    Irradiation-induced recrystallization appears to be a general phenomenon in that it is observed to occur in a variety of nuclear fuel types, e.g. U-xMo, UO2, and U3O8. For temperatures below that where significant thermal annealing of defects occurs, an expression is derived for the fission density at which irradiation-induced recrystallization is initiated that is athermal and weakly dependent on fission rate. The initiation of recrystallization is to be distinguished from the subsequent progression and eventual consumption of the original fuel grain. The formulation takes into account the observed microstructural evolution of the fuel, the role of precipitate pinning and fission gas bubbles, and the triggering event for recrystallization. The calculated dislocation density, fission gas bubble-size distribution, and fission density at which recrystallization first appears are compared to measured quantities.

  11. Fission product release and microstructure changes of irradiated MOX fuel at high temperatures

    NASA Astrophysics Data System (ADS)

    Colle, J.-Y.; Hiernaut, J.-P.; Wiss, T.; Beneš, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

    2013-11-01

    Samples of irradiated MOX fuel of 44.5 GWd/tHM mean burn-up were prepared by core drilling at three different radial positions of a fuel pellet. They were subsequently heated in a Knudsen effusion mass spectrometer up to complete vaporisation of the sample (˜2600 K) and the release of fission gas (krypton and xenon) as well as helium was measured. Scanning electron microscopy was used in parallel to investigate the evolution of the microstructure of a sample heated under the same condition up to given key temperatures as determined from the gas release profiles. A clear initial difference for fission gas release and microstructure was observed as a function of the radial position of the samples and therefore of irradiation temperature. A good correlation between the microstructure evolution and the gas release peaks could be established as a function of the temperature of irradiation and (laboratory) heating. The region closest to the cladding (0.58 < r/r0 < 0.96), designated as sample type A in Fig. 1. It represents the "cooler" part of the fuel pellet. The irradiation temperatures (Tirrad) in this range are from 854 to 1312 K (ΔT: 458 K). The intermediate radial zone of the pellet (0.42 < r/r0 < 0.81), designated sample type B in Fig. 1, has a Tirrad ranging from 1068 to 1434 K (ΔT: 365 K). The central zone of the pellet (0.003 < r/r0 < 0.41), designated sample type C in Fig. 1, which was close to the hottest part of the pellet, has a Tirrad ranging from 1442 to 1572 K (ΔT: 131 K). The sample irradiation temperatures were determined from the calculated temperature profile (exponential function) knowing the core temperature of the fuel (1573 K) [11], the standard temperature for this type of fuel at the inner side of the cladding (800 K). The average burnup was calculated with TRANSURANUS code [12] and the PA burnup is the average burnup multiplied by the ratio of the fissile Pu concentration in PA over average fissile Pu concentration in fuel [11]. Calculated

  12. A scoping study of fission product transport from failed fuel during N Reactor postulated accidents

    SciTech Connect

    Hagrman, D.L.

    1988-01-01

    This report presents a scoping study of cesium, iodine, and tellurium behavior during a cold leg manifold break in the N Reactor. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The LACE LA1 test and evidence from the Power Burst Facility Severe Fuel Damage tests are used to test the key model applied to determine aerosol behavior. Recommendations for future analysis are also provided. The primary result is that most of the cesium, iodine, and tellurium remains in the molten uranium fuel. Only 0.0035 of the total inventory is calculated to be released. Condensation of the most of the species of cesium and iodine that are released is calculated, with 0.998 of the released cesium and iodine condensing in the spacers and upstream end of the connector tubes. Most of the tellurium that is released condenses, but the chemical reaction of tellurium vapor with surfaces is also a major factor in the behavior of this element.

  13. Fission product retention in TRISCO coated UO sub 2 particle fuels subjected to HTR simulated core heating tests

    SciTech Connect

    Baldwin, C.A.; Kania, M.J.

    1990-11-01

    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600{degree}C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800{degree}C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800{degree}C and above may exist. 6 refs., 6 figs., 4 tabs.

  14. Analysis of Neutron Fission Reaction Rate in the Nuclear Fuel Cell Using Collision Probability Method with Non Flat Flux Approach

    NASA Astrophysics Data System (ADS)

    Shafii, Mohammad Ali

    2017-07-01

    Neutron fission reaction rate in the nuclear reactor depends on macroscopic cross section and neutron flux distribution. The macroscopic cross section depends on the type of nuclide, the type of reaction, and the group energy of the neutrons relative to the nuclides. Flux distribution is very important in a nuclear reactor, because it is closely related to power distribution. In general, the integral neutron transport equation is solved using a collision probability (CP) method with a flat flux (FF) approach. Consequently, the CP matrix is also assumed constantly, therefore, the distribution of the neutron flux throughout the cell becomes flat. In the non-flat flux (NFF) approach, the neutron flux is modellled by linear interpolation as a function of mesh in the cylindrical nuclear fuel cell of a fast reactor type. This study uses the CP method with a NFF approach and it is applied to analyze the neutron fission reaction rate of a cylindrical nuclear fuel cell of a fast reactor type. Nuclear data library that is used in this study is JFS-3-J33 which belongs to the SLAROM computer code. Calculation results of the fission reaction rate shows that it is decrease in the high energy region due to the events of elastic collision that caused the neutron easier to lose of energy. The same fission reaction rate pattern occurs in the FF and NFF approaches.

  15. In-pile release behavior of metallic fission products in graphite materials of an htgr fuel assembly

    NASA Astrophysics Data System (ADS)

    Hayashi, K.; Kobayashi, F.; Minato, K.; Ikawa, K.; Fukuda, K.

    1987-06-01

    Distribution of metallic fission products in the graphite sleeve and block of the fifth OGL-1 fuel assembly was measured by gamma spectrometry with lathe sectioning. Considerably large release fractions of long-lived fission products with smooth axial profiles were observed in the sleeve due to a large failure fraction of coated fuel particles accompanied with failed silicon carbide layers. Nevertheless, a key nuclide 110mAg, whose large release is suspected at increased burnups for low-enriched uranium fuels, was effectively retained within the graphite sleeve. The retention was also observed for 125Sb, 154Eu and 155Eu up to a burnup of 3.2% fission per initial metal atom, but was limited for 134Cs and 137Cs at high sleeve-temperatures above 900°C. In-pile diffusion coefficients in IG-110 graphite have been evaluated for Cs, Ag and Sb; those for Cs are in reasonable agreement with available in-pile data.

  16. 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.

  17. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 <20 pct) U-Mo dispersion fuel is being developed for use in research and test reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  18. A separate effect study of the influence of metallic fission products on CsI radioactive release from nuclear fuel

    NASA Astrophysics Data System (ADS)

    Di Lemma, F. G.; Colle, J. Y.; Beneš, O.; Konings, R. J. M.

    2015-10-01

    The chemistry of cesium and iodine is of main importance to quantify the radioactive release in case of a nuclear reactor accident, or sabotage involving irradiated nuclear materials. We studied the interaction of CsI with different metallic fission products such as Mo and Ru. These elements can be released from nuclear fuel when exposed to oxidising conditions, as in the case of contact of overheated nuclear fuel with air (e.g. in a spent fuel cask sabotage, uncovering of a spent fuel pond, or air ingress accidents). Experiments were performed by vaporizing mixtures of the compounds in air, and analysing the produced aerosols in view of a possible gas-gas and gas-aerosol reactions between the compounds. These results were compared with the gaseous species predicted by thermochemical equilibrium calculations and experimental equilibrium vaporization tests using Knudsen Effusion Mass Spectrometry.

  19. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  20. Gamma-ray spectrometric measurements of fission rate ratios between fresh and burnt fuel following irradiation in a zero-power reactor

    NASA Astrophysics Data System (ADS)

    Kröhnert, H.; Perret, G.; Murphy, M. F.; Chawla, R.

    2013-01-01

    The gamma-ray activity from short-lived fission products has been measured in fresh and burnt UO2 fuel samples after irradiation in a zero-power reactor. For the first time, short-lived gamma-ray activity from fresh and burnt fuel has been compared and fresh-to-burnt fuel fission rate ratios have been derived. For the measurements, well characterized fresh and burnt fuel samples, with burn-ups up to 46 GWd/t, were irradiated in the zero-power research reactor PROTEUS. Fission rate ratios were derived based on the counting of high-energy gamma-rays above 2200 keV, in order to discriminate against the high intrinsic activity of the burnt fuel. This paper presents the measured fresh-to-burnt fuel fission rate ratios based on the 142La (2542 keV), 89Rb (2570 keV), 138Cs (2640 keV) and 95Y (3576 keV) high-energy gamma-ray lines. Comparisons are made with the results of Monte Carlo modeling of the experimental configuration, carried out using the MCNPX code. The measured fission rate ratios have 1σ uncertainties of 1.7-3.4%. The comparisons with calculated predictions show an agreement within 1-3σ, although there appears to be a slight bias (∼3%).

  1. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    NASA Astrophysics Data System (ADS)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

  2. Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air

    SciTech Connect

    Trumbull, TH

    2004-10-18

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

  3. An initial assessment of a mechanistic model, GRASS-SST, in U-Pu-Zr metallic alloy fuel fission-gas behavior simulations

    NASA Astrophysics Data System (ADS)

    Yun, Di; Rest, Jeffrey; Hofman, Gerard L.; Yacout, Abdellatif M.

    2013-04-01

    A mechanistic kinetic rate theory model originally developed for the prediction of fission gas behavior in oxide nuclear fuels under steady-state and transient conditions has been assessed to investigate its applicability to model fission gas behavior in U-Pu-Zr metallic alloy fuel. In order to capture and validate the underlying physics for irradiated U-Pu-Zr fuels, the mechanistic model was applied to evaluate fission gas release, fission gas and fission product induced swelling, and detailed gas bubble size distributions in three different fuel zones: the outer α-U, the intermediate, and the inner γ-U zones. Due to its special microstructural features, the α-U zone in U-Pu-Zr fuels is believed to contribute the largest fraction of fission gas release among the different fuel zones. It is shown that with the use of small effective grain sizes, the mechanistic model can predict fission gas release that is in reasonable consistence with (though slightly lower than) experimentally measured data. These simulation results are comparable to the experimentally measured fission gas release since the mechanism of fission gas transport through the densely distributed laminar porosity in the α-U zone is analogous to the mechanism of fission gas transport through the interconnected gas bubble porosity utilized in the mechanistic model. Detailed gas bubble size distributions predicted with the mechanistic model in both the intermediate zone and the high temperature γ-U zone of U-Pu-Zr fuel are also compared to experimental measurements from available SEM micrographs. These comparisons show good agreement between the simulation results and experimental measurements, and therefore provide crucial guidelines for the selection of key physical parameters required for modeling these two zones. Material properties such as fuel grain size and thermal diffusivity of gas and model parameters such as di-atom nucleation probability and gas bubble re-solution constant are predicted

  4. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    SciTech Connect

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

  5. Test irradiations of full-sized U 3Si 2-Al fuel plates up to very high fission densities

    NASA Astrophysics Data System (ADS)

    Böning, K.; Petry, W.

    2009-01-01

    In the course of the licensing procedure of the 'Forschungsneutronenquelle Heinz Maier-Leibnitz', i.e. the new 20 MW high-flux research reactor FRM II in Garching near Munich, extensive test irradiations have been performed to qualify the U 3Si 2-Al dispersion fuel with a relatively high density of highly enriched uranium (93 wt% of 235U) up to very high fission densities. Two of the three FRM II type fuel plates used in the irradiation tests contained U 3Si 2-Al dispersion fuel with HEU densities of 3.0 gU/cm 3 or 1.5 gU/cm 3 ('homogeneous plates') and one plate had two adjacent zones of either density ('mixed plate'). They were irradiated in the French MTR reactors SILOE and OSIRIS in the years before 2002. The local plate thickness was measured on certain tracks along the plates during interruptions of the irradiation. The maximum fission density obtained in the U 3Si 2 fuel particles was 1.4 × 10 22 f/cm 3 and 1.1 × 10 22 f/cm 3 in the 1.5 gU/cm 3 and 3.0 gU/cm 3 fuel zones, respectively. In the course of the irradiations, the plate thickness increased monotonically and approximately linearly, leading to a maximum plate thickness swelling of 14% and 21% and a corresponding volume increase of the fuel particles of 106% and 81%, respectively. Our results are discussed and compared with the data from the literature.

  6. Licensing topical report: the measurement and modelling of time-dependent fission product release from failed HTGR fuel particles under accident conditions

    SciTech Connect

    Myers, B.F.; Morrissey, R.E.

    1980-04-01

    The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission products was measured from laser-failed BISO ThO/sub 2/ and highly enriched (HEU) TRISO UC/sub 2/ particles that had been irradiated to a range of kernel burnups. The burnups were 0.25, 1.4, and 15.7% FIMA for ThO/sub 2/ particles and 23.5 and 74% FIMA for UC/sub 2/ particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium.

  7. Fission yields data generation and benchmarks of decay heat estimation of a nuclear fuel

    NASA Astrophysics Data System (ADS)

    Gil, Choong-Sup; Kim, Do Heon; Yoo, Jae Kwon; Lee, Jounghwa

    2017-09-01

    Fission yields data with the ENDF-6 format of 235U, 239Pu, and several actinides dependent on incident neutron energies have been generated using the GEF code. In addition, fission yields data libraries of ORIGEN-S, -ARP modules in the SCALE code, have been generated with the new data. The decay heats by ORIGEN-S using the new fission yields data have been calculated and compared with the measured data for validation in this study. The fission yields data ORIGEN-S libraries based on ENDF/B-VII.1, JEFF-3.1.1, and JENDL/FPY-2011 have also been generated, and decay heats were calculated using the ORIGEN-S libraries for analyses and comparisons.

  8. Using gamma spectrometry indicators to detect and quantify fission products changes in irradiated fuel

    SciTech Connect

    Loubet, L.; Martella, Th.

    2015-07-01

    A new analysis method based on gamma scanning of fission products on irradiated rods is presented. Indicators calculated from this method can be used for the qualitative treatment and comparison of irradiated rods from PWR, SFR or and MTR. Differences in the behavior of fission products (FP) can thus be quantified. Phenomena such as migration or geometrical changes in pellets should thus benefit from these accurate, yet quickly and easily achievable results. (authors)

  9. Effects of radiation and fission product incorporation in a yttria-stabilized zirconia based inert matrix fuel

    NASA Astrophysics Data System (ADS)

    Zhu, Sha

    This work has investigated the irradiation and incorporation effects of fission products in a yttria-stabilized zirconia (YSZ) based inert matrix fuel (IMF). The concept of inert matrix fuel is based on a new strategy for disposition of plutonium generated from the reprocessing of commercial nuclear fuel and the dismantling of nuclear weapons, i.e. using uranium-free oxides to "burn" plutonium and other actinides (Np, Cm, and Am) in reactors. This approach allows direct disposal, without reprocessing, after once-through burn-up. YSZ and MgAl2O4-YSZ composites are among the potential ceramics for IMF due to their high chemical durability and radiation resistance. The research involved investigating the production, nature, and accumulation of irradiation-induced defects, the behavior of the fission products in the ceramics, the structural stability and amorphization resistance of the YSZ during implantation. Ion implantations were conducted with 200--400 keV Cs+, Sr+, I+, Xe+ and Ti+ up to fluences of 1 x 1017/cm 2 at both room temperature and temperatures of 600--700°C. Thermal annealing was subsequently completed after room temperature ion implantations. In situ and ex situ transmission electron microscopy (TEM), optical absorption spectroscopy, photo-luminescence spectroscopy, and electron paramagnetic resonance (EPR) spectroscopy were employed to characterize the irradiation induced defect evolution and analyze the defect structures. Various irradiation effects were observed and determined in the experiments, such as point defects (F type and V type color centers), defect clusters (dislocation loops), cavities (voids and bubbles), the crystalline-to-amorphous transition, and the phase transformation from fluorite to pyrochlore structure. The ion irradiation-induced amorphization mechanism, the retention ability of the fission products, and structural stability of YSZ are discussed in terms of ion incorporation effects, implanted ion radii, and the solubility

  10. Numerical Study on Effects of Fuel Mixture Fraction and Li-6 Enrichment on Neutronic Parameters of a Fusion-Fission Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Yapııcıı, Hüseyin; Genç, Gamze; Demir, Nesrin

    2004-09-01

    This study presents the effects of mixture fractions of nuclear fuels (mixture of fissile-fertile fuels and mixture of two different fertile fuels) and 6Li enrichment on the neutronic parameters (the tritium breeding ratio, TBR, the fission rate, FR, the energy multiplication ratio, M, the fissile breeding rate, FBR, the neutron leakage out of blanket, L, and the peak-to-average fission power density ratio, Γ) of a deuterium-tritium (D-T) fusion neutron-driven hybrid blanket. Three different fertile fuels (232Th, 238U and 244Cm), and one fissile fuel (235U) were selected as the nuclear fuel. Two different coolants (pressurized helium and natural lithium) were used for the nuclear heat transfer out of the fuel zone (FZ). The Boltzmann transport equation was solved numerically for obtaining the neutronic parameters with the help of the neutron transport code XSDRNPM/SCALE4.4a. In addition, these calculations were performed by also using the MCNP4B code. The sub-limits of the mixture fractions and 6Li enrichment were determined for the tritium self-sufficiency. The considered hybrid reactor can be operated in a self-sufficiency mode in the cases with the fuel mixtures mixed with a fraction of equal to or greater than these sub-limits. Furthermore, the numerical results show that the fissile fuel breeding and fission potentials of the blankets with the helium coolant are higher than with the lithium coolant.

  11. Quantification of Uncertainties due to 235,238U, 239,240,241Pu and Fission Products Nuclear Data Uncertainties for a PWR Fuel Assembly

    NASA Astrophysics Data System (ADS)

    da Cruz, D. F.; Rochman, D.; Koning, A. J.

    2014-04-01

    Uncertainty analysis on reactivity and discharged inventory for a typical PWR fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. The Total Monte-Carlo (TMC) method was applied using the deterministic transport code DRAGON. The nuclear data used in this study is from the JEFF-3.1 evaluations, with the exception of the nuclear data files for U, Pu and fission products isotopes, which are taken from the nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as result of uncertainties in nuclear data of the considered isotopes) is virtually independent on fuel burnp and amounts to 700 pcm. The uncertainties in inventory of the discharged fuel is dependent on the element considered and lies in the range 1-15% for most fission products, and is below 5% for the most important actinides.

  12. Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets

    NASA Astrophysics Data System (ADS)

    Zhao, Yunmei; Ding, Shurong; Zhang, Xunchao; Wang, Canglong; Yang, Lei

    2016-12-01

    The micro-scale finite element models for CERCER pellets with different-sized fuel particles are developed. With consideration of a grain-scale mechanistic irradiation swelling model in the fuel particles and the irradiation creep in the matrix, numerical simulations are performed to explore the effects of the particle size and the fission-fragment-enhanced irradiation creep on the thermo-mechanical behavior of CERCER pellets. The enhanced irradiation creep effect is applied in the 10 μm-thick fission fragment damage matrix layer surrounding the fuel particles. The obtained results indicate that (1) lower maximum temperature occurs in the cases with smaller-sized particles, and the effects of particle size on the mechanical behavior in pellets are intricate; (2) the first principal stress and radial axial stress remain compressive in the fission fragment damage layer at higher burnup, thus the mechanism of radial cracking found in the experiment can be better explained.

  13. Determining fissile content in PWR spent fuel assemblies using a passive neutron Albedo reactivity with fission chambers technique

    SciTech Connect

    Conlin, Jeremy Lloyd; Tobin, Stephen J

    2010-01-01

    State regulatory bodies and organizations such as the IAEA that are concerned with preventing the proliferation of nuclear weapons are interested in a means of quantifying the amount of plutonium in a given spent fuel assembly. The complexity of spent nuclear fuel makes the measurement of plutonium content challenging. There are a variety of techniques that can measure various properties of spent nuclear fuel including burnup, and mass of fissile content. No single technique can provide all desired information, necessitating an approach using multiple detector systems and types. This paper presents our analysis of the Passive Neutron Albedo Reactivity Fission Chamber (PNAR-FC) detector system. PNAR-FC is a simplified version of the PNAR technique originally developed in 1997. This earlier research was performed with a high efficiency, {sup 3}He-based system (PNAR-3He) with which multiplicty analysis was performed. With the PNAR technique a portion of the spent fuel assembly is wrapped in a 1 mm thick cadmium liner. Neutron count rates are measured both with and without the cadmium liner present. The ratio of the count rate with the cadmium liner to the count rate without the cadmium liner is calculated and called the cadmium ratio. In the PNAR-3He technique, multiplicity measurements were made and the cadmium ratio was shown to scale with the fissile content of the material being measured. PNAR-FC simplifies the PNAR technique by using only a few fission chambers instead of many {sup 3}He tubes. Using a simplified PNAR-FC technique provides for a cheaper, lighter, and thus more portable detector system than was possible with the PNAR-3He system. The challenge with the PNAR-FC system are two-fold: (1) the change in the cadmium ratio is weaker as a afunction of the changing fissile content relative to multiplicity count rates, and (2) the efficiency for the fission chamber based system are poorer than for the {sup 3}He based detectors. In this paper, we present our

  14. Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients

    NASA Astrophysics Data System (ADS)

    Iqbal, M. Javed; Mirza, Nasir M.; Mirza, Sikander M.

    2008-01-01

    During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135I, 135Xe and 89Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.

  15. Freshly induced short-lived gamma-ray activity as a measure of fission rates in lightly re-irradiated spent fuel

    NASA Astrophysics Data System (ADS)

    Kröhnert, H.; Perret, G.; Murphy, M. F.; Chawla, R.

    2010-12-01

    A new measurement technique has been developed to determine fission rates in burnt fuel, following re-irradiation in a zero-power research reactor. The development has been made in the frame of the LIFE@PROTEUS program at the Paul Scherrer Institute, which aims at characterizing the interfaces between fresh and highly burnt fuel assemblies in modern LWRs. To discriminate against the high intrinsic gamma-ray activity of the burnt fuel, the proposed measurement technique uses high-energy gamma-rays, above 2000 keV, emitted by short-lived fission products freshly produced in the fuel. To demonstrate the feasibility of this technique, a fresh UO 2 sample and a 36 GWd/t burnt UO 2 sample were irradiated in the PROTEUS reactor and their gamma-ray activities were recorded directly after irradiation. For both fresh and the burnt fuel samples, relative fission rates were derived for different core positions, based on the short-lived 142La (2542 keV), 89Rb (2570 keV), 138Cs (2640 keV) and 95Y (3576 keV) gamma-ray lines. Uncertainties on the inter-position fission rate ratios were mainly due to the uncertainties on the net-area of the gamma-ray peaks and were about 1-3% for the fresh sample, and 3-6% for the burnt one. Thus, for the first time, it has been shown that the short-lived gamma-ray activity, induced in burnt fuel by irradiation in a zero-power reactor, can be used as a quantitative measure of the fission rate. For both fresh and burnt fuel, the measured results agreed, within the uncertainties, with Monte Carlo (MCNPX) predictions.

  16. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    NASA Astrophysics Data System (ADS)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  17. Experimental validation of CASMO-4E and CASMO-5M for radial fission rate distributions in a westinghouse SVEA-96 Optima2 BWR fuel assembly

    SciTech Connect

    Grimm, P.; Perret, G.

    2012-07-01

    Measured and calculated radial total fission rate distributions are compared for the three axial sections of a Westinghouse SVEA-96 Optima2 BWR fuel assembly, comprising 96, 92 and 84 fuel rods, respectively. The measurements were performed on a full-size fuel assembly in the PROTEUS zero-power experimental facility. The measured fission rates are compared to the results of the CASMO-4E and CASMO-5M fuel assembly codes. Detailed measured geometrical data were used in the models, and effects of the surrounding zones of the reactor were taken into account by correction factors derived from MCNPX calculations. The results of the calculations agree well with those of the experiments, with root-mean-square deviations between 1.2% and 1.5% and maximum deviations of 3-4%. The quality of the predictions by CASMO-4E and CASMO-5M is comparable. (authors)

  18. Volatile fission product behaviour during thermal annealing of irradiated UO 2 fuel oxidised up to U 3O 8

    NASA Astrophysics Data System (ADS)

    Hiernaut, J.-P.; Wiss, T.; Papaioannou, D.; Konings, R. J. M.; Rondinella, V. V.

    2008-01-01

    The behaviour and release of fission products during high-temperature annealing of irradiated UO 2 samples have been studied as a function of the oxidation state. The behaviour of a sample pre-oxidised to U 3O 8 was compared to that of non-pre-treated fuel from the same pellet radial location. The Knudsen cell mass spectrometer technique was used up to 1900 K for the pre-oxidised sample and up to 2800 K for the untreated sample. Both types of tests were run in vacuum. The possible chemical forms of the different fission products in the bulk and in the vapour phase have been estimated from the release curves and microprobe analysis. This study concerns essentially iodine, tellurium, caesium, rubidium, strontium, barium, technetium and molybdenum, whose effusion behaviour was strongly affected by the pre-oxidation treatment, resulting in an almost complete release by 1900 K. Release of zirconium, the lanthanides and actinides was observed at temperatures >1900 K, reached only in the case of the non-pre-treated UO 2 experiments.

  19. Advances in Development of the Fission Product Extraction Process for the Separation of Cesium and Strontium from Spent Nuclear Fuel

    SciTech Connect

    JAck D. Law

    2007-09-01

    The Fission Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Advanced Fuel Cycle Initiative for the simultaneous separation of cesium (Cs) and strontium (Sr) from spent light water reactor (LWR) fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository, and when combined with the separation of americium (Am) and curium (Cm), could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with a simulated feed solution in 3.3-cm centrifugal contactors are detailed. Removal efficiencies, distribution coefficient data, coextraction of metals, and process hydrodynamic performance are discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel.

  20. I-NERI ANNUAL TECHNICAL PROGRESS REPORT: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels

    SciTech Connect

    S. Frank

    2009-09-01

    An attractive alternative to the once-through disposal of electrorefiner salt is to selectively remove the active fission products from the salt and recycle the salt back to the electrorefiner (ER). This would allow salt reuse for some number of cycles before ultimate disposal of the salt in a ceramic waste form. Reuse of ER salt would, thus, greatly reduce the volume of ceramic waste produced during the pyroprocessing of spent nuclear fuel. This final portion of the joint I-NERI research project is to demonstrate the separation of fission products from molten ER salt by two methods previously selected during phase two (FY-08) of this project. The two methods selected were salt/zeolite contacting and rare-earth fission product precipitation by oxygen bubbling. The ER salt used in these tests came from the Mark-IV electrorefiner used to anodically dissolved driver fuel from the EBR-II reactor on the INL site. The tests were performed using the Hot Fuel Dissolution Apparatus (HFDA) located in the main cell of the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels complex on the INL site. Results from these tests were evaluated during a joint meeting of KAERI and INL investigators to provide recommendations as to the future direction of fission product removal from electrorefiner salt that accumulate during spent fuel treatment. Additionally, work continued on kinetic measurements of surrogate quaternary salt systems to provide fundamental kinetics on the ion exchange system and to expand the equilibrium model system developed during the first two phases of this project. The specific objectives of the FY09 I-NERI research activities at the INL include the following: • Perform demonstration tests of the selected KAERI precipitation and INL salt/zeolite contacting processes for fission product removal using radioactive, fission product loaded ER salt • Continue kinetic studies of the quaternary Cs/Sr-LiCl-KCl system to determine the rate of ion

  1. HETEROGENEOUS NUCLEAR REACTOR EMPLOYING SMALL UNCLAD BODIES OF FISSIONABLE MATERIAL AS FUEL

    DOEpatents

    Hyman, H.H.; Katz, J.J.

    1961-05-01

    A nuclear reactor in which fuel pellets are continuously dissolved in a moderator liquid is described. The fuel pellets are fed into the top of elongated baskets which are submerged in moderator liquid, and a portion of the moderator liquid is continuously withdrawn and processed to recove r reaction products.

  2. Fuel and fission product behaviour in early phases of a severe accident. Part I: Experimental results of the PHEBUS FPT2 test

    NASA Astrophysics Data System (ADS)

    Barrachin, M.; Gavillet, D.; Dubourg, R.; De Bremaecker, A.

    2014-10-01

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO2 fuel test section and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 and 900 mm) of the 1-m long test section are presented in this paper. Material interactions leading to local corium formation were identified: firstly between fuel and Zircaloy-4 cladding, notably at 823 mm, where the cladding melting temperature was reached, and secondly between fuel and stainless steel oxides. Regarding fission products, molybdenum left so-called metallic precipitates mainly composed of ruthenium. Xenon and caesium behave similarly whereas barium and molybdenum often seems to be associated in precipitates.

  3. Fission gas detection system

    DOEpatents

    Colburn, Richard P.

    1985-01-01

    A device for collecting fission gas released by a failed fuel rod which device uses a filter to pass coolant but which filter blocks fission gas bubbles which cannot pass through the filter due to the surface tension of the bubble.

  4. Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel

    NASA Astrophysics Data System (ADS)

    Lillo, T. M.; van Rooyen, I. J.

    2016-05-01

    In this study, the fission product precipitates at silicon carbide grain boundaries from an irradiated TRISO particle were identified and correlated with the associated grain boundary characteristics. Precession electron diffraction in the transmission electron microscope provided the crystallographic information needed to identify grain boundary misorientation and boundary type (i.e., low angle, random high angle or coincident site lattice (CSL)-related). The silicon carbide layer was found to be composed mainly of twin boundaries and small fractions of random high angle and low angle grain boundaries. Most fission products were found at random, high-angle grain boundaries, with small fractions at low-angle and CSL-related grain boundaries. Palladium (Pd) was found at all types of grain boundaries while Pd-uranium and Pd-silver precipitates were only associated with CSL-related and random, high-angle grain boundaries. Precipitates containing only Ag were found only at random, high-angle grain boundaries, but not at low angle or CSL-related grain boundaries.

  5. TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

    SciTech Connect

    Powers, Jeffrey James

    2011-11-30

    This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

  6. Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992--May 31, 1995

    SciTech Connect

    Sforza, P.M.; Cresci, R.J.

    1997-05-31

    Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an innovative application of a base injection stabilized recirculation bubble is presented. The development of the experimental facility, a simulated thrust chamber approximately 0.4 m in diameter and 1 m long, is described. The flow rate of propellant simulant (air) can be varied up to about 2 kg/sec and that of fuel simulant (air, air-sulfur hexafluoride) up to about 0.2 kg/sec. This scale leads to chamber Reynolds numbers on the same order of magnitude as those anticipated in a full-scale nuclear rocket engine. The experimental program introduced here is focused on determining the size, geometry, and stability of the recirculation region as a function of the bleed ratio, i.e. the ratio of the injected mass flux to the free stream mass flux. A concurrent CFD study is being carried out to aid in demonstrating that the proposed technique is practical.

  7. Neutronic evaluation of fissile fuel breeding blankets for the fission-suppressed Tandem-Mirror Hybrid Reactor

    SciTech Connect

    Johnson, J.O.; Burns, T.J.

    1984-06-01

    A computational study was performed on the blanket design of the Lawrence Livermore National Laboratory (LLNL) fission-suppressed Tandem Mirror Hybrid Reactor (TMHR) to qualify the methods and data bases available at Oak Ridge National Laboratory (ORNL) for use in analyzing the neutronic performance of fissile fuel breeding blankets. The eventual goal of the study was to establish the capability for analysis and optimization of advanced fissile fuel production blanket designs. Discrete ordinates radiation transport calculations were performed in one-dimensional cylindrical geometry to obtain the blanket spatial distribution and energy spectra of the neutron and gamma-ray fluxes resulting from the monoenergetic (14.1 MeV) fusion first wall source. Key macroscopic cross sections of the blanket materials were then folded with the flux spectra to obtain reaction rates critical to evaluating blanket feasibility. Finally, a time-dependent depletion analysis was performed to evaluate the blanket performance during equilibrium cycle conditions. The results of the study are presented both as graphs and tables.

  8. Combustion of HTR fuel: Fission product release and dissolution of plutonium contained in the ash

    NASA Astrophysics Data System (ADS)

    Zimmer, E.; Brodda, B. G.; Kirchner, H.; Kroth, K.

    1985-04-01

    Single spent spherical HTR fuel elements were subjected to combustion between 800 and 1100°C. The elements were loaded with either UO 2 or (Th, U)O 2 pyrocarbon coated fuel kernels. The UO 2 loaded elements released a large fraction of their cesium and ruthenium inventory during combustion, whereas the release of the (Th, U)O 2 loaded element was significantly lower. The ash of the UO 2 elements was dissolved in nitric acid to simulate Purex reprocessing. About 5% of the bred plutonium was indissoluble and remained in the residue. The spent (Th, U)O 2 fuel was broken into pieces after combustion and dissolved five times faster than fresh fuel in fluoride containing nitric acid.

  9. DEVELOPMENT OF GLASS COMPOSITIONS TO IMMOBILIZE ALKALI, ALKALINE EARTH, LANTHANIDE AND TRANSITION METAL FISSION PRODUCTS FROM NUCLEAR FUEL REPROCESSING

    SciTech Connect

    Marra, J.; Billings, A.

    2009-06-24

    The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product (TM) wastes is vitrification into a borosilicate glass. A current interest is to evaluate the feasibility of vitrifying combined waste streams to most cost effectively immobilize the wastes resulting from aqueous fuel reprocessing. Studies showed that high waste loadings are achievable for the Ln only (Option 1) stream. Waste loadings in excess of 60 wt % (on a calcined oxide basis) were demonstrated via a lanthanide borosilicate (LaBS) glass. The resulting glasses had excellent relative durability as determined by the Product Consistency Test (PCT). For a combined Ln and TM waste stream glass (Option 2), noble metal solubility was found to limit waste loading. However, the measured PCT normalized elemental releases for this glass were at least an order of magnitude below that of Environmental Assessment (EA) glass. Current efforts to evaluate the feasibility of vitrifying combined Ln, TM, alkali (Cs is the primary radionuclide of concern) and alkaline earth (Sr is the primary radionuclide of concern) wastes (Option 3) have shown that these approaches are feasible. However, waste loading limitations with respect to heat load (Cs/Sr loading), molybdenum solubility and/or noble metal solubility will likely be realized and must be considered in determining the cost effectiveness of these approaches.

  10. Nano-Scale Fission Product Phases in an Irradiated U-7Mo Alloy Nuclear Fuel

    SciTech Connect

    Dennis Keiser, Jr.; Brandon Miller; James Madden; Jan-Fong Jue; Jian Gan

    2014-09-01

    Irradiated nuclear fuel is a very difficult material to characterize. Due to the large radiation fields associated with these materials, they are hard to handle and typically have to be contained in large hot cells. Even the equipment used for performing characterization is housed in hot cells or shielded glove boxes. The result is not only a limitation in the techniques that can be employed for characterization, but also a limitation in the size of features that can be resolved The most standard characterization techniques include light optical metallography (WM), scanning electron microscopy (SEM), and electron probe microanalysis (EPMA). These techniques are applied to samples that are typically prepared using grinding and polishing approaches that will always generate some mechanical damage on the sample surface. As a result, when performing SEM analysis, for example, the analysis is limited by the quality of the sample surface that can be prepared. However, a new approach for characterizing irradiated nuclear fuel has recently been developed at the Idaho National Laboratory (INL) in Idaho Falls, Idaho. It allows for a dramatic improvement in the quality of characterization that can be performed when using an instrument like an SEM. This new approach uses a dual-beam scanning microscope, where one of the beams isa focused ion beam (FIB), which can be used to generate specimens of irradiated fuel (-10µm x 10µm) for microstructural characterization, and the other beam is the electron beam of an SEM. One significant benefit of this approach is that the specimen surface being characterized has received much less damage (and smearing) than is caused by the more traditional approaches, which enables the imaging of nanometer­ sized microstructural features in the SEM. The process details are for an irradiated low-enriched uranium (LEU) U-Mo alloy fuel Another type of irradiated fuel that has been characterized using this technique is a mixed oxide fuel.

  11. A complementary approach to estimate the internal pressure of fission gas bubbles by SEM-SIMS-EPMA in irradiated nuclear fuels

    NASA Astrophysics Data System (ADS)

    Cagna, C.; Zacharie-Aubrun, I.; Bienvenu, P.; Barrallier, L.; Michel, B.; Noirot, J.

    2016-02-01

    The behaviour of gases produced by fission is of great importance for nuclear fuel in operation. Within this context, a decade ago, a general method for the characterisation of the fission gas including gas bubbles in an irradiated UO2 nuclear fuel was developed and applied to determine the bubbles internal pressure. The method consists in the determination of the pressure, over a large population of bubbles, using three techniques: SEM, EPMA and SIMS. In this paper, a complementary approach using the information given by the same techniques is performed on an isolated bubble under the surface and is aiming for a better accuracy compared to the more general measurement of gas content. SEM and EPMA enable the detection of a bubble filled with xenon under the surface. SIMS enables the detection of the gas filling the bubble. The quantification is achieved using the EPMA data as reference at positions where no or nearly no bubbles are detected.

  12. The DART dispersion analysis research tool: A mechanistic model for predicting fission-product-induced swelling of aluminum dispersion fuels. User`s guide for mainframe, workstation, and personal computer applications

    SciTech Connect

    Rest, J.

    1995-08-01

    This report describes the primary physical models that form the basis of the DART mechanistic computer model for calculating fission-product-induced swelling of aluminum dispersion fuels; the calculated results are compared with test data. In addition, DART calculates irradiation-induced changes in the thermal conductivity of the dispersion fuel, as well as fuel restructuring due to aluminum fuel reaction, amorphization, and recrystallization. Input instructions for execution on mainframe, workstation, and personal computers are provided, as is a description of DART output. The theory of fission gas behavior and its effect on fuel swelling is discussed. The behavior of these fission products in both crystalline and amorphous fuel and in the presence of irradiation-induced recrystallization and crystalline-to-amorphous-phase change phenomena is presented, as are models for these irradiation-induced processes.

  13. Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution

    DOEpatents

    Tomczuk, Zygmunt; Miller, William E.; Wolson, Raymond D.; Gay, Eddie C.

    1991-01-01

    An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U.sup.+4 cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd.sup.+2 cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode.

  14. FEMAXI-V benchmarking study on peak temperature and fission gas release prediction of PWR rod fuel

    NASA Astrophysics Data System (ADS)

    Suwardi, Dewayatna, W.; Briyatmoko, B.

    2012-06-01

    The present paper reports a study of FEMAXI-V code and related report on code benchmarking. Capabilities of the FEMAXI-V code to predict the thermal and fission gas release have been tested on MOX fuels in LWRs which has been done in SCK.CEN & Belgonucleaire by using PRIMO MOX rod BD8 irradiation experiment after V Sobolev as reported O. J. Ott [2]. Base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor (ramp power excursion). The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop, installed on a movable structure of the core periphery. The power variations were obtained by inwards/backwards movements of the loop in the core water. The preconditioning phase for rod BD8 occurred at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/ (cm.min), reaching a terminal peak power level of 395 W/cm that lasted for 20 hours.

  15. FEMAXI-V benchmarking study on peak temperature and fission gas release prediction of PWR rod fuel

    SciTech Connect

    Suwardi; Dewayatna, W.; Briyatmoko, B.

    2012-06-06

    The present paper reports a study of FEMAXI-V code and related report on code benchmarking. Capabilities of the FEMAXI-V code to predict the thermal and fission gas release have been tested on MOX fuels in LWRs which has been done in SCK{center_dot}CEN and Belgonucleaire by using PRIMO MOX rod BD8 irradiation experiment after V Sobolev as reported O. J. Ott. Base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor (ramp power excursion). The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop, installed on a movable structure of the core periphery. The power variations were obtained by inwards/backwards movements of the loop in the core water. The preconditioning phase for rod BD8 occurred at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/ (cm.min), reaching a terminal peak power level of 395 W/cm that lasted for 20 hours.

  16. Chemical states of fission products in irradiated (U 0.3Pu 0.7)C 1+ x fuel at high burn-ups

    NASA Astrophysics Data System (ADS)

    Agarwal, Renu; Venugopal, V.

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel, (U 0.3Pu 0.7)C 1+ x, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  17. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2±x: Implications for nuclear fuel performance modeling

    NASA Astrophysics Data System (ADS)

    Andersson, D. A.; Garcia, P.; Liu, X.-Y.; Pastore, G.; Tonks, M.; Millett, P.; Dorado, B.; Gaston, D. R.; Andrs, D.; Williamson, R. L.; Martineau, R. C.; Uberuaga, B. P.; Stanek, C. R.

    2014-08-01

    Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2±x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2±x non-stoichiometry were used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2±x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Risø fuel rod irradiation experiment was simulated.

  18. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2 +/- x: Implications for nuclear fuel performance modeling

    SciTech Connect

    Giovanni Pastore; Michael R. Tonks; Derek R. Gaston; Richard L. Williamson; David Andrs; Richard Martineau

    2014-03-01

    Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2x nonstoichiometrywere used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Risø fuel rod irradiation experiment was simulated. 2014 Elsevier B.V. All rights

  19. Kinetic Monte Carlo (KMC) simulation of fission product silver transport through TRISO fuel particle

    NASA Astrophysics Data System (ADS)

    de Bellefon, G. M.; Wirth, B. D.

    2011-06-01

    A mesoscale kinetic Monte Carlo (KMC) model developed to investigate the diffusion of silver through the pyrolytic carbon and silicon carbide containment layers of a TRISO fuel particle is described. The release of radioactive silver from TRISO particles has been studied for nearly three decades, yet the mechanisms governing silver transport are not fully understood. This model atomically resolves Ag, but provides a mesoscale medium of carbon and silicon carbide, which can include a variety of defects including grain boundaries, reflective interfaces, cracks, and radiation-induced cavities that can either accelerate silver diffusion or slow diffusion by acting as traps for silver. The key input parameters to the model (diffusion coefficients, trap binding energies, interface characteristics) are determined from available experimental data, or parametrically varied, until more precise values become available from lower length scale modeling or experiment. The predicted results, in terms of the time/temperature dependence of silver release during post-irradiation annealing and the variability of silver release from particle to particle have been compared to available experimental data from the German HTR Fuel Program ( Gontard and Nabielek [1]) and Minato and co-workers ( Minato et al. [2]).

  20. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    NASA Astrophysics Data System (ADS)

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-11-01

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy's Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

  1. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    DOE PAGES

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; ...

    2016-08-15

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representativemore » of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.« less

  2. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    SciTech Connect

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-08-15

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

  3. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    SciTech Connect

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-08-15

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

  4. Computer program FPIP-REV calculates fission product inventory for U-235 fission

    NASA Technical Reports Server (NTRS)

    Brown, W. S.; Call, D. W.

    1967-01-01

    Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

  5. Fuel and fission product behaviour in early phases of a severe accident. Part II: Interpretation of the experimental results of the PHEBUS FPT2 test

    NASA Astrophysics Data System (ADS)

    Dubourg, R.; Barrachin, M.; Ducher, R.; Gavillet, D.; De Bremaecker, A.

    2014-10-01

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO2 fuel bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

  6. Thermo-Mechanical Response of a TRISO Fuel Particle in a Fusion/Fission Engine for Incineration of Weapons Grade Plutonium

    SciTech Connect

    Caro, M; DeMange, P; Marian, J; Caro, A

    2009-12-08

    The Laser Inertial Fusion-based (LIFE) engine is an advanced energy concept under development at Lawrence Livermore National Laboratory (LLNL). LIFE engine could be used to drive a subcritical fission blanket with fertile or fissile fuel. Current LIFE engine designs envisages fuel in pebble bed form with TRISO (tristructural isotropic) particles embedded in a graphite matrix, and pebbles flowing in molten salt Flibe (2LiF+BeF{sub 2}) coolant at T {approx} 700C. Weapons-grade plutonium (WGPu) fuel is an attractive option for LIFE engine involving the achievement of high fractional burnups in a short lifetime frame. However, WGPu LIFE engine operating conditions of high neutron fast fluence, high radiation damage, and high Helium and Hydrogen production pose severe challenges for typical TRISO particles. The thermo-mechanical fuel performance code HUPPCO (High burn-Up fuel Pebble Performance COde) currently under development accounts for spatial and time dependence of the material elastic properties, temperature, and irradiation swelling and creep mechanisms. In this work, some aspects of the thermo-mechanical response of TRISO particles used for incineration of weapons grade fuel in LIFE engine are analyzed. Preliminary results show the importance of developing reliable high-fidelity models of the performance of these new fuel designs and the need of new experimental data relevant to WGPu LIFE conditions.

  7. Gas-leaking fuel elements number and fission gas product coolant volumetric activities assessment in the VVER-440 nuclear power plant

    NASA Astrophysics Data System (ADS)

    Szuta, Marcin

    1992-07-01

    In a nuclear power plant it is required to monitor continuously the number of gas-leaking fuel elements and the contamination level of the primary coolant by fission gas products. It is proposed to use the radiation monitoring system equipped with the computer technics provided with the suitable program package for fulfilment this requirements. The input data to start up the program consists of the 88Kr volumetric activity measured by the radiation monitoring system and three actual technological parameters: coolant temperature at inlet, thermal power and coolant flow rate.

  8. Determination of the rod-wise fission gas release fraction in a complete fuel assembly using non-destructive gamma emission tomography

    NASA Astrophysics Data System (ADS)

    Holcombe, Scott; Andersson, Peter; Svärd, Staffan Jacobsson; Hallstadius, Lars

    2016-11-01

    A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel. In this work, the tomographic device has been experimentally benchmarked for non-destructive rod-wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel-stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of 85Kr in the gas plenum to 137Cs in the fuel stack. The rod-wise ratio of 85Kr/137Cs was, in turn, used to calculate the rod-wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method. Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO2, and four rods were from an assembly with a burnup of 26 MWd/kgUO2. At the time of the measurements, the nine rods had cooled for

  9. USA/FRG umbrella agreement for cooperation in GCR [Gas Cooled Reactor] development: Fuel, fission products and graphite subprogram. Part 1, Management meeting report: Part 2, Revised subprogram plan, Revision 10

    SciTech Connect

    1986-05-01

    This Subprogram Plan describes cooperative work in the areas of HTR fuel and graphite development and fission product studies that is being carried out under US/FRG/Swiss Implementing Agreement for cooperation in Gas Cooled Reactor development. Only bilateral US/FRG cooperation is included, since it is the only active work in this subprogram area at this time. The cooperation has been in progress since February 1977. A number of Project Work Statements have been developed in each of the major areas of the subprogram, and work on many of them is in progress. The following specific areas are included in the scope of this plan: fuel development; graphite development; fission product release; and fission product behavior outside the fuel elements.

  10. Neutronic Model of a Mirror Based Fusion-Fission Hybrid for the Incineration of Spent Nuclear Fuel and with Potential for Energy Amplification

    NASA Astrophysics Data System (ADS)

    Noack, Klaus; Moiseenko, V. E.; Agren, O.; Hagnestall, A.

    2010-11-01

    In the last decade the Georgia Institute of Technology (Georgia Tech) published several design concepts of tokamak based fusion-fission hybrids which use solid fuels consisting of transuranic elements of the spent nuclear fuel from Light-Water-Reactors. The objectives of the hybrids are the incineration of the transuranic elements and an additional net energy production under the condition of tritium self-sufficiency. The present paper presents a preliminary scientific design of the blanket of a mirror based hybrid which was derived from the results of Monte Carlo neutron transport calculations. The main operation parameters of two hybrid options were specified. One is the analog to Georgia Techs first version of a ``fusion transmutation of waste reactor'' (FTWR) and the other is a possible near-term option which requires minimal fusion power. The latter version shows considerably better performance parameters.

  11. The Testing of Recent JEF(F) Decay Data and Fission Product Yields Files for Irradiated Nuclear Fuel Decay Heat Calculations

    NASA Astrophysics Data System (ADS)

    Mills, R. W.; Parker, D. R.

    2005-05-01

    The heat generated by irradiated nuclear fuel is one of the important considerations for its safe storage, transport and possible recycling. One method to calculate the decay heat of irradiated fuel is from an inventory code such as FISPIN or ORIGEN-S. These codes were part of a code comparison that showed that if using the same nuclear data their results for a set of testcases differed by less than 1 part in 103. This paper compares FISPIN decay heat calculations with a selection of fission pulse experiments (U235, U238, Pu239 and Pu241) and UOX PWR assembly calorimetric measurements. The calculations were performed using libraries based upon JEF-1 (1986), JEF-2.2 (1993) and a preliminary JEFF-3 file that includes a UK fission product yield file (UKFY3.5). The results show that both JEF-2.2 and the preliminary JEFF-3 data predict the decay heat to a similar accuracy and generally within 5%.

  12. Thermodynamics of fission products in dispersion fuel designs - first principles modeling of defect behavior in bulk and at interfaces

    SciTech Connect

    Liu, Xiang-yand; Uberuaga, Blas P; Nerikar, Pankaj; Sickafus, Kurt E; Stanek, Chris R

    2009-01-01

    Density functional theory (DFT) calculations of fission product (Xe, Sr, and Cs) incorporation and segregation in alkaline earth metal oxides, HfO{sub 2} and UO{sub 2} oxides, and the MgO/(U, Hf, Ce)O{sub 2} interfaces have been carried out. In the case of UO{sub 2}, the calculations were performed using spin polarization and with a Hubbard U term characterizing the on-sit Coulomb repulsion between the localized 5f electrons. The fission product solution energies in bulk UO{sub 2{+-}x} have been calculated as a function of non-stoichiometry x, and were compared to that in MgO. These calculations demonstrate that the fission product incorporation energies in MgO are higher than in HfO{sub 2}. However, this trend is reversed or reduced for alkaline earth oxides with larger cation sizes. The solution energies of fission products in MgO are substantially higher than in UO{sub 2{+-}x}, except for the case of Sr in the hypostoichiometric case. Due to size effects, the thermodynamic driving force of segregation for Xe and Cs from bulk MgO to the MgO/fluorite interface is strong. However, this driving force is relatively weak for Sr.

  13. Nuclear Fission

    NASA Astrophysics Data System (ADS)

    Denschlag, J. O.

    This chapter first gives a survey on the history of the discovery of nuclear fission. It briefly presents the liquid-drop and shell models and their application to the fission process. The most important quantities accessible to experimental determination such as mass yields, nuclear charge distribution, prompt neutron emission, kinetic energy distribution, ternary fragment yields, angular distributions, and properties of fission isomers are presented as well as the instrumentation and techniques used for their measurement. The contribution concentrates on the fundamental aspects of nuclear fission. The practical aspects of nuclear fission are discussed in http://dx.doi.org/10.1007/978-1-4419-0720-2_57 of Vol. 6.

  14. Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

    NASA Astrophysics Data System (ADS)

    Noirot, J.; Lamontagne, J.; Nakae, N.; Kitagawa, T.; Kosaka, Y.; Tverberg, T.

    2013-11-01

    A UO2 fuel with a heterogeneous distribution of 235U was irradiated up to a high burn-up in the Halden Boiling Water Reactor (HBWR). The last 100 days of irradiation were performed with an increased level of linear power. The effect of the heterogeneous fissile isotope distribution on the formation of the HBS was studied free of the possible influence of Pu which exists in heterogeneous MOX fuels. The HBS formed in 235U-rich agglomerates and its main characteristics were very similar to those of the HBS formed in Pu-rich agglomerates of heterogeneous MOX fuels. The maximum local contents of Nd and Xe before HBS formation were studied in this fuel. In addition to a Pu effect that promotes the HBS phenomenon, comparison with previous results for heterogeneous MOX fuels showed that the local fission product concentration was not the only parameter that has to be taken into consideration. It appears that the local actinide depletion by fission and/or the energy locally deposited through electronic interactions in the fission fragment recoils also have an effect on the HBS formation threshold. Moreover, a major release of fission gases from the peripheral 235U-rich agglomerates of HBS bubbles and a Cs radial movement are also evidenced in this heterogeneous UO2. Cs deposits on the peripheral grain boundaries, including the HBS grain boundaries, are considered to reveal the release paths. SUP>235U-rich agglomerates, SUP>235U-poor areas, an intermediate phase with intermediate 235U concentrations. Short fuel rods were fabricated with these pellets. The main characteristics of these fuel rods are shown in Table 1.These rods were irradiated to high burn-ups in the IFA-609/626 of the HBWR and then one was irradiated in the IFA-702 for 100 days. Fig. 2 shows the irradiation history of this fuel. The final average burn-up of the rod was 69 GWd/tU. Due to the flux differences along the rod, however, the average burn-up of the cross section examined was 63 GWd/tU. This fuel

  15. Nuclear Power from Fission Reactors. An Introduction.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  16. Use of fission track analysis technique for the determination of MicroBequerel level of 239Pu in urine samples from radiation workers handling MOX fuel.

    PubMed

    Yadav, J R; Rao, D D; Kumar, Ranjeet; Aggarwal, S K

    2011-07-01

    Fission track analysis (FTA) technique for the determination of (239)Pu excreted through urine has been standardized using blank samples, tracer and (239)Pu spikes. Double stage anion exchange separation protocol has been applied and an average radiochemical recovery of (239)Pu of 18% was obtained. An average track registration efficiency of 11 tracks per μBq of (239)Pu, irradiated to 0.35×10(17) neutron fluence was established. Reagent blank urine samples from 11 controlled subjects were analyzed by FTA and an average of 149±14 tracks was obtained. Minimum detectable activity of 34μBqL(-1) of urine sample was obtained and will be useful for monitoring chronic exposure cases handling MOX fuel.

  17. Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

    2015-12-01

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  18. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    SciTech Connect

    Shmelev, A. N. Kulikov, G. G. Kurnaev, V. A. Salahutdinov, G. H. Kulikov, E. G. Apse, V. A.

    2015-12-15

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  19. Spontaneous Fission

    DOE R&D Accomplishments Database

    Segre, Emilio

    1950-11-22

    The first attempt to discover spontaneous fission in uranium was made by [Willard] Libby, who, however, failed to detect it on account of the smallness of effect. In 1940, [K. A.] Petrzhak and [G. N.] Flerov, using more sensitive methods, discovered spontaneous fission in uranium and gave some rough estimates of the spontaneous fission decay constant of this substance. Subsequently, extensive experimental work on the subject has been performed by several investigators and will be quoted in the various sections. [N.] Bohr and [A.] Wheeler have given a theory of the effect based on the usual ideas of penetration of potential barriers. On this project spontaneous fission has been studied for the past several years in an effort to obtain a complete picture of the phenomenon. For this purpose the spontaneous fission decay constants {lambda} have been measured for separated isotopes of the heavy elements wherever possible. Moreover, the number {nu} of neutrons emitted per fission has been measured wherever feasible, and other characteristics of the spontaneous fission process have been studied. This report summarizes the spontaneous fission work done at Los Alamos up to January 1, 1945. A chronological record of the work is contained in the Los Alamos monthly reports.

  20. Nuclear Fission Research at IRMM

    SciTech Connect

    Hambsch, Franz-Josef

    2005-05-24

    The Institute for Reference Materials and Measurements (IRMM) will celebrate its 45th anniversary in 2005. With its 150-MeV Geel Electron Linear Accelerator (GELINA) and 7-MV Van de Graaff accelerator as multi-purpose neutron sources, it served the nuclear physics community for this period.The research in the field of nuclear fission was focused in recent years on both the measurement and calculation of fission cross sections, and the measurement of fission fragment properties.Fission cross sections were determined for 233Pa and 234U; the fission process was studied in the resolved resonance region of 239Pu(n,f) and for 251Cf(nth,f). These measurements derive their interest from accelerator driven systems, the thorium fuel cycle, high temperature reactors, safety issues of current reactors, and basic physics. The measurements are supported by several modeling efforts that aim at improving model codes and nuclear data evaluation.

  1. The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors: A preliminary assessment of experiments HRB-17, HFR-B1, HFR-K6 and KORA

    SciTech Connect

    Myers, B.F.

    1995-09-01

    The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors has been measured in different laboratories under both irradiation and post irradiation conditions. The data from experiments HRB-17, HFR-B1, HFR-K6, and in the KORA facility are compared to assess their consistency and complimentarily. The experiments are consistent under comparable experimental conditions and reveal two general mechanisms involving exposed fuel kernels embedded in carbonaceous materials. One is manifest as a strong dependence of fission gas release on the partial pressure of water vapor below 1 kPa and the other, as a weak dependence above 1 kPa.

  2. Analysis of fission and activation radionuclides produced by a uranium-fueled nuclear detonation and identification of the top dose-producing radionuclides.

    PubMed

    Kraus, Terry; Foster, Kevin

    2014-08-01

    The radiological assessment of the nuclear fallout (i.e., fission and neutron-activation radionuclides) from a nuclear detonation is complicated by the large number of fallout radionuclides. This paper provides the initial isotopic source term inventory of the fallout from a uranium-fueled nuclear detonation and identifies the significant and insignificant radiological dose producing radionuclides over 11 dose integration time periods (time phases) of interest. A primary goal of this work is to produce a set of consistent, time phase-dependent lists of the top dose-producing radionuclides that can be used to prepare radiological assessment calculations and data products (e.g., maps of areas that exceed protective action guidelines) in support of public and worker protection decisions. The ranked lists of top dose-producing radionuclides enable assessors to perform atmospheric dispersion modeling and radiological dose assessment modeling more quickly by using relatively short lists of radionuclides without significantly compromising the accuracy of the modeling and the dose projections. This paper also provides a superset-list of the top dose-producing fallout radionuclides from a uranium-fueled nuclear detonation that can be used to perform radiological assessments over any desired time phase. Furthermore, this paper provides information that may be useful to monitoring and sampling and laboratory analysis personnel to help understand which radionuclides are of primary concern. Finally, this paper may be useful to public protection decision makers because it shows the importance of quickly initiating public protection actions to minimize the radiological dose from fallout.

  3. DESIGN OF AN ON-LINE, MULTI-SPECTROMETER FISSION PRODUCT MONITORING SYSTEM (FPMS) TO SUPPORT ADVANCED GAS REACTOR (AGR) FUEL TESTING AND QUALIFICATION IN THE ADVANCED TEST REACTOR

    SciTech Connect

    J. K. Hartwell; D. M. Scates; M. W. Drigert

    2005-11-01

    The US Department of Energy (DOE) is embarking on a series of tests of coated-particle reactor fuel for the Advanced Gas Reactor (AGR). As one part of this fuel development program a series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). The first test in this series (AGR-1) will incorporate six separate “capsules” irradiated simultaneously, each containing about 51,000 TRISO-coated fuel particles supported in a graphite matrix and continuously swept with inert gas during irradiation. The effluent gas from each of the six capsules must be independently monitored in near real time and the activity of various fission gas nuclides determined and reported. A set of seven heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based total radiation detectors have been designed, and are being configured and tested for use during the AGR-1 experiment. The AGR-1 test specification requires that the AGR-1 fission product measurement system (FPMS) have sufficient sensitivity to detect the failure of a single coated fuel particle and sufficient range to allow it to “count” multiple (up to 250) successive particle failures. This paper describes the design and expected performance of the AGR-1 FPMS.

  4. Advanced Space Fission Propulsion Systems

    NASA Technical Reports Server (NTRS)

    Houts, Michael G.; Borowski, Stanley K.

    2010-01-01

    Fission has been considered for in-space propulsion since the 1940s. Nuclear Thermal Propulsion (NTP) systems underwent extensive development from 1955-1973, completing 20 full power ground tests and achieving specific impulses nearly twice that of the best chemical propulsion systems. Space fission power systems (which may eventually enable Nuclear Electric Propulsion) have been flown in space by both the United States and the Former Soviet Union. Fission is the most developed and understood of the nuclear propulsion options (e.g. fission, fusion, antimatter, etc.), and fission has enjoyed tremendous terrestrial success for nearly 7 decades. Current space nuclear research and technology efforts are focused on devising and developing first generation systems that are safe, reliable and affordable. For propulsion, the focus is on nuclear thermal rockets that build on technologies and systems developed and tested under the Rover/NERVA and related programs from the Apollo era. NTP Affordability is achieved through use of previously developed fuels and materials, modern analytical techniques and test strategies, and development of a small engine for ground and flight technology demonstration. Initial NTP systems will be capable of achieving an Isp of 900 s at a relatively high thrust-to-weight ratio. The development and use of first generation space fission power and propulsion systems will provide new, game changing capabilities for NASA. In addition, development and use of these systems will provide the foundation for developing extremely advanced power and propulsion systems capable of routinely and affordably accessing any point in the solar system. The energy density of fissile fuel (8 x 10(exp 13) Joules/kg) is more than adequate for enabling extensive exploration and utilization of the solar system. For space fission propulsion systems, the key is converting the virtually unlimited energy of fission into thrust at the desired specific impulse and thrust

  5. Radiochemistry and the Study of Fission

    SciTech Connect

    Rundberg, Robert S.

    2016-11-14

    These are slides from a lecture given at UC Berkeley. Radiochemistry has been used to study fission since it’ discovery. Radiochemical methods are used to determine cumulative mass yields. These measurements have led to the two-mode fission hypothesis to model the neutron energy dependence of fission product yields. Fission product yields can be used for the nuclear forensics of nuclear explosions. The mass yield curve depends on both the fuel and the neutron spectrum of a device. Recent studies have shown that the nuclear structure of the compound nucleus can affect the mass yield distribution. The following topics are covered: In the beginning: the discovery of fission; forensics using fission products: what can be learned from fission products, definitions of R-values and Q-values, fission bases, K-factors and fission chambers, limitations; the neutron energy dependence of the mass yield distribution (the two mode fission hypothesis); the influence of nuclear structure on the mass yield distribution. In summary: Radiochemistry has been used to study fission since it’s discovery. Radiochemical measurement of fission product yields have provided the highest precision data for developing fission models and for nuclear forensics. The two-mode fission hypothesis provides a description of the neutron energy dependence of the mass yield curve. However, data is still rather sparse and more work is needed near second and third chance fission. Radiochemical measurements have provided evidence for the importance of nuclear states in the compound nucleus in predicting the mass yield curve in the resonance region.

  6. Space Fission Propulsion System Development Status

    NASA Technical Reports Server (NTRS)

    Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Williams, Eric; Harper, Roger; Salvail, Pat; Hrbud, Ivana; hide

    2001-01-01

    The world's first man-made self-sustaining fission reaction was achieved in 1942. Since then fission has been used to propel submarines, generate tremendous amounts of electricity, produce medical isotopes, and provide numerous other benefits to society. Fission systems operate independently of solar proximity or orientation, and are thus well suited for deep spare or planetary surface missions. In addition, the fuel for fission systems (enriched uranium) is virtually non-radioactive. The primary safety issue with fission systems is avoiding inadvertent system start - addressing this issue through proper system design is straightforward. Despite the relative simplicity and tremendous potential of space fission systems, the development and utilization of these systems has proven elusive. The first use of fission technology in space occurred 3 April 1965 with the US launch of the SNAP-10A reactor. There have been no additional US uses of space fission system. While space fission system were used extensively by the former Soviet Union, their application was limited to earth-orbital missions. Early space fission systems must be safely and affordably utilized if Ae are to reap the benefits of advanced space fission systems.

  7. Space Fission Propulsion System Development Status

    NASA Technical Reports Server (NTRS)

    Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Williams, Eric; Harper, Roger; Salvail, Pat; Hrbud, Ivana; Rodgers, Stephen L. (Technical Monitor)

    2001-01-01

    The world's first man-made self-sustaining fission reaction was achieved in 1942. Since then fission has been used to propel submarines, generate tremendous amounts of electricity, produce medical isotopes, and provide numerous other benefits to society. Fission systems operate independently of solar proximity or orientation, and are thus well suited for deep spare or planetary surface missions. In addition, the fuel for fission systems (enriched uranium) is virtually non-radioactive. The primary safety issue with fission systems is avoiding inadvertent system start - addressing this issue through proper system design is straightforward. Despite the relative simplicity and tremendous potential of space fission systems, the development and utilization of these systems has proven elusive. The first use of fission technology in space occurred 3 April 1965 with the US launch of the SNAP-10A reactor. There have been no additional US uses of space fission system. While space fission system were used extensively by the former Soviet Union, their application was limited to earth-orbital missions. Early space fission systems must be safely and affordably utilized if Ae are to reap the benefits of advanced space fission systems.

  8. Benchmarking nuclear fission theory

    DOE PAGES

    Bertsch, G. F.; Loveland, W.; Nazarewicz, W.; ...

    2015-05-14

    We suggest a small set of fission observables to be used as test cases for validation of theoretical calculations. Thus, the purpose is to provide common data to facilitate the comparison of different fission theories and models. The proposed observables are chosen from fission barriers, spontaneous fission lifetimes, fission yield characteristics, and fission isomer excitation energies.

  9. Benchmarking nuclear fission theory

    SciTech Connect

    Bertsch, G. F.; Loveland, W.; Nazarewicz, W.; Talou, P.

    2015-05-14

    We suggest a small set of fission observables to be used as test cases for validation of theoretical calculations. Thus, the purpose is to provide common data to facilitate the comparison of different fission theories and models. The proposed observables are chosen from fission barriers, spontaneous fission lifetimes, fission yield characteristics, and fission isomer excitation energies.

  10. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-10-03

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  11. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-01-01

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  12. Fuel pin

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  13. Fuel pin

    SciTech Connect

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  14. Neutronics for critical fission reactors and subcritical fission in hybrids

    SciTech Connect

    Salvatores, Massimo

    2012-06-19

    The requirements of future innovative nuclear fuel cycles will focus on safety, sustainability and radioactive waste minimization. Critical fast neutron reactors and sub-critical, external source driven systems (accelerator driven and fusion-fission hybrids) have a potential role to meet these requirements in view of their physics characteristics. This paper provides a short introduction to these features.

  15. A hemi-fission intermediate links two mechanistically distinct stages of membrane fission

    PubMed Central

    Sundborger, Anna C.; Hortelano, Eva Rodriguez; Fuhrmans, Marc; Neumann, Sylvia; Müller, Marcus; Hinshaw, Jenny E.; Schmid, Sandra L.; Frolov, Vadim A.

    2015-01-01

    Fusion and fission drive all vesicular transport. Although topologically opposite, these reactions pass through the same hemi-fusion/fission intermediate1,2, characterized by a ‘stalk’ in which only the inner monolayers of the two compartments have merged to form a localized non-bilayer connection1-3. Formation of the hemi-fission intermediate requires energy input from proteins catalyzing membrane remodeling; however the relationship between protein conformational rearrangements and hemi-fusion/fission remains obscure. Here we analyzed how the GTPase cycle of dynamin, the prototypical membrane fission catalyst4-6, is directly coupled to membrane remodeling. We used intra-molecular chemical cross-linking to stabilize dynamin in its GDP•AlF4--bound transition-state. In the absence of GTP this conformer produced stable hemi-fission, but failed to progress to complete fission, even in the presence of GTP. Further analysis revealed that the pleckstrin homology domain (PHD) locked in its membrane-inserted state facilitated hemi-fission. A second mode of dynamin activity, fueled by GTP hydrolysis, couples dynamin disassembly with cooperative diminishing of the PHD wedging, thus destabilizing the hemi-fission intermediate to complete fission. Molecular simulations corroborate the bimodal character of dynamin action and indicate radial and axial forces as dominant, although not independent drivers of hemi-fission and fission transformations, respectively. Mirrored in the fusion reaction7-8, the force bimodality might constitute a general paradigm for leakage-free membrane remodeling. PMID:26123023

  16. Fission Spectrum

    DOE R&D Accomplishments Database

    Bloch, F.; Staub, H.

    1943-08-18

    Measurements of the spectrum of the fission neutrons of 25 are described, in which the energy of the neutrons is determined from the ionization produced by individual hydrogen recoils. The slow neutrons producing fission are obtained by slowing down the fast neutrons from the Be-D reaction of the Stanford cyclotron. In order to distinguish between fission neutrons and the remaining fast cyclotron neutrons both the cyclotron current and the pusle amplifier are modulated. A hollow neutron container, in which slow neutrons have a lifetime of about 2 milliseconds, avoids the use of large distances. This method results in much higher intensities than the usual modulation arrangement. The results show a continuous distribution of neutrons with a rather wide maximum at about 0.8 MV falling off to half of its maximum value at 2.0 MV. The total number of netrons is determined by comparison with the number of fission fragments. The result seems to indicate that only about 30% of the neutrons have energies below .8 MV. Various tests are described which were performed in order to rule out modification of the spectrum by inelastic scattering. Decl. May 4, 1951

  17. Bimodal fission

    SciTech Connect

    Hulet, E.K.

    1989-04-19

    In recent years, we have measured the mass and kinetic-energy distributions from the spontaneous fission of /sup 258/Fm, /sup 259/Md, /sup 260/Md, /sup 258/No, /sup 262/No, and /sup 260/(104). All are observed to fission with a symmetrical division of mass, whereas the total-kinetic-energy (TKE) distributions strongly deviated from the Gaussian shape characteristically found in the fission of all other actinides. When the TKE distributions are resolved into two Gaussians the constituent peaks lie near 200 and near 233 MeV. We conclude two modes or bimodal fission is occurring in five of the six nuclides studied. Both modes are possible in the same nuclides, but one generally predominates. We also conclude the low-energy but mass-symmetrical mode is likely to extend to far heavier nuclei; while the high-energy mode will be restricted to a smaller region, a region of nuclei defined by the proximity of the fragments to the strong neutron and proton shells in /sup 132/Sn. 16 refs., 7 figs., 1 tab.

  18. Space Fission Propulsion System Development Status

    NASA Technical Reports Server (NTRS)

    Houts, M.; Van Dyke, M. K.; Godfroy, T. J.; Pedersen, K. W.; Martin, J. J.; Dickens, R.; Williams, E.; Harper, R.; Salvail, P.; Hrbud, I.

    2001-01-01

    The world's first man-made self-sustaining fission reaction was achieved in 1942. Since then fission has been used to propel submarines, generate tremendous amounts of electricity, produce medical isotopes, and provide numerous other benefits to society. Fission systems operate independently of solar proximity or orientation, and are thus well suited for deep space or planetary surface missions. In addition, the fuel for fission systems (enriched uranium) is virtually non-radioactive. The primary safety issue with fission systems is avoiding inadvertent system start. Addressing this issue through proper system design is straight-forward. Despite the relative simplicity and tremendous potential of space fission systems, the development and utilization of these systems has proven elusive. The first use of fission technology in space occurred 3 April 1965 with the US launch of the SNAP-10A reactor. There have been no additional US uses of space fission systems. While space fission systems were used extensively by the former Soviet Union, their application was limited to earth-orbital missions. Early space fission systems must be safely and affordably utilized if we are to reap the benefits of advanced space fission systems. NASA's Marshall Space Flight Center, working with Los Alamos National Laboratory (LANL), Sandia National Laboratories, and others, has conducted preliminary research related to a Safe Affordable Fission Engine (SAFE). An unfueled core has been fabricated by LANL, and resistance heaters used to verify predicted core thermal performance by closely mimicking heat from fission. The core is designed to use only established nuclear technology and be highly testable. In FY01 an energy conversion system and thruster will be coupled to the core, resulting in an 'end-to-end' nuclear electric propulsion demonstrator being tested using resistance heaters to closely mimic heat from fission. Results of the SAFE test program will be presented. The applicability

  19. Space Fission Propulsion System Development Status

    NASA Technical Reports Server (NTRS)

    Houts, M.; Van Dyke, M. K.; Godfroy, T. J.; Pedersen, K. W.; Martin, J. J.; Dickens, R.; Williams, E.; Harper, R.; Salvail, P.; Hrbud, I.

    2001-01-01

    The world's first man-made self-sustaining fission reaction was achieved in 1942. Since then fission has been used to propel submarines, generate tremendous amounts of electricity, produce medical isotopes, and provide numerous other benefits to society. Fission systems operate independently of solar proximity or orientation, and are thus well suited for deep space or planetary surface missions. In addition, the fuel for fission systems (enriched uranium) is virtually non-radioactive. The primary safety issue with fission systems is avoiding inadvertent system start. Addressing this issue through proper system design is straight-forward. Despite the relative simplicity and tremendous potential of space fission systems, the development and utilization of these systems has proven elusive. The first use of fission technology in space occurred 3 April 1965 with the US launch of the SNAP-10A reactor. There have been no additional US uses of space fission systems. While space fission systems were used extensively by the former Soviet Union, their application was limited to earth-orbital missions. Early space fission systems must be safely and affordably utilized if we are to reap the benefits of advanced space fission systems. NASA's Marshall Space Flight Center, working with Los Alamos National Laboratory (LANL), Sandia National Laboratories, and others, has conducted preliminary research related to a Safe Affordable Fission Engine (SAFE). An unfueled core has been fabricated by LANL, and resistance heaters used to verify predicted core thermal performance by closely mimicking heat from fission. The core is designed to use only established nuclear technology and be highly testable. In FY01 an energy conversion system and thruster will be coupled to the core, resulting in an 'end-to-end' nuclear electric propulsion demonstrator being tested using resistance heaters to closely mimic heat from fission. Results of the SAFE test program will be presented. The applicability

  20. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  1. Fission meter

    DOEpatents

    Rowland, Mark S [Alamo, CA; Snyderman, Neal J [Berkeley, CA

    2012-04-10

    A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source.

  2. Distribution of fission products in an LMFBR: summary

    SciTech Connect

    Villarreal, R.; Young, J.O.

    1985-01-01

    The overall distribution of fission products released from experimental subassemblies containing breached fuel elements has been determined in the fuel and throughout the EBR-II primary and secondary reactor systems. Identification of the fission products released to the primary sodium and location of areas of concentration was important in anticipating radioactive species and levels of deposited fission and activation products on components removed from the primary tank for maintenance and repair. The results of extensive radioanalytical measurements on the fuel, fuel cladding, primary sodium and cover gas system, secondary sodium and cover gas system and steam system are summarized.

  3. FUEL ELEMENT

    DOEpatents

    Fortescue, P.; Zumwalt, L.R.

    1961-11-28

    A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

  4. Fission-suppressed hybrid reactor: the fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  5. Liquid uranium alloy-helium fission reactor

    DOEpatents

    Minkov, Vladimir

    1986-01-01

    This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

  6. Thermodynamic analysis of volatile organometallic fission products.

    PubMed

    Auxier, John D; Jordan, Jacob A; Stratz, S Adam; Shahbazi, Shayan; Hanson, Daniel E; Cressy, Derek; Hall, Howard L

    The ability to perform rapid separations in a post nuclear weapon detonation scenario is an important aspect of national security. In the past, separations of fission products have been performed using solvent extraction, precipitation, etc. The focus of this work is to explore the feasibility of using thermochromatography, a technique largely employed in superheavy element chemistry, to expedite the separation of fission products from fuel components. A series of fission product complexes were synthesized and the thermodynamic parameters were measured using TGA/DSC methods. Once measured, these parameters were used to predict their retention times using thermochromatography.

  7. SABR Fusion-Fission Hybrid Studies

    NASA Astrophysics Data System (ADS)

    Stewart, Chris

    2012-03-01

    The Subcritical Advanced Burner Reactor (SABR) concept is a fast reactor comprised of a tokamak fusion neutron source based on ITER surrounded by an annular fission core adapted from Integral Fast Reactor designs. Previous work has examined SABR used to help close the nuclear fuel cycle by fissioning the transuranics from spent nuclear fuel. One focus of the present work is a SABR Breeder Reactor to achieve tritium self-sufficieny and a Pu breeding ratio significantly above 1 in order to provide fuel for SABR as well as for MOX-fueled LWR's and other fast reactors. Another focus of this research is the dynamic safety simulation of lloss-of-flow loss-of-heat-sink, loss-of-power, and positive reactivity accidents in the TRU fuel SABR burner reactor. The reactivity effect of thermal-induced bowing of fuel pins has been modeled, which is expected to provide passive safety.

  8. FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Abbott, W.E.; Balent, R.

    1958-09-16

    A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

  9. Microstructural Characterization of Irradiated U-7Mo/Al-5Si Dispersion to High Fission Density

    SciTech Connect

    J. Gan; B. D. Miller; D. D. Keiser, Jr.; A. B. Robinson; J. W. Madden; P. G. Medvedev; D. M. Wachs

    2014-11-01

    The fuel development program for research and test reactors calls for improved knowledge on the effect of microstructure on fuel performance in reactors. This work summarizes the recent TEM microstructural characterization of an irradiated U-7Mo/Al-5Si dispersion fuel plate (R3R050) irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory to 5.2×1021 fissions/cm3. While a large fraction of the fuel grains is decorated with large bubbles, there is no evidence showing interlinking of these large bubbles at the specified fission density. The attachment of solid fission product precipitates to the bubbles is likely the result of fission product diffusion into these bubbles. The process of fission gas bubble superlattice collapse appears through bubble coalescence. The results are compared with the previous TEM work of the dispersion fuels irradiated to lower fission density from the same fuel plate.

  10. Italian hybrid and fission reactors scenario analysis

    SciTech Connect

    Ciotti, M.; Manzano, J.; Sepielli, M.

    2012-06-19

    Italy is a country where a long tradition of studies both in the fission and fusion field is consolidated; nevertheless a strong public opinion concerned with the destination of the Spent Nuclear Fuel hinders the development of nuclear power. The possibility to a severe reduction of the NSF mass generated from a fleet of nuclear reactors employing an hypothetical fusionfission hybrid reactor has been investigated in the Italian framework. The possibility to produce nuclear fuel for the fission nuclear reactors with the hybrid reactor was analyzed too.

  11. Italian hybrid and fission reactors scenario analysis

    NASA Astrophysics Data System (ADS)

    Ciotti, M.; Manzano, J.; Sepielli, M.

    2012-06-01

    Italy is a country where a long tradition of studies both in the fission and fusion field is consolidated; nevertheless a strong public opinion concerned with the destination of the Spent Nuclear Fuel hinders the development of nuclear power. The possibility to a severe reduction of the NSF mass generated from a fleet of nuclear reactors employing an hypothetical fusionfission hybrid reactor has been investigated in the Italian framework. The possibility to produce nuclear fuel for the fission nuclear reactors with the hybrid reactor was analyzed too.

  12. Marmot-Fission-Gas-Diffusion

    SciTech Connect

    Andersson, Anders; Matthews, Christopher

    2016-10-22

    The MARMOT-FISSION-GAS-DIFFUSION software solves a coupled set of partial differential equations describing fission gas evolution in UO2 nuclear fuel. It is part of the MARMOT code, which builds on the MOOSE framework. Both the MARMOT code and the MOOSE framework are developed and maintained by Idaho National Laboratory. The model in MARMOT-FISSION-GAS-DIFFUSION consists of a set of continuum reaction-diffusion equations capturing formation and annihilation of defects, reactions between defects, diffusion of defects and segregation of defects to grain boundaries. Defects refer to vacancies and interstitials as well fission gas atoms (Xe) occupying various trap sites such as uranium and oxygen vacancies and interstitials sites. The code can treat a large number of defect types. The model is formulated within the phase field framework to be compatible with other MARMOT kernels. The driving forces for all reactions, diffusion and segregation events are consistently formulated as a variational derivatives of the free energy of the system. The rates of the reactions are controlled by the corresponding kinetic coefficients. The free energy and the kinetic coefficients for UO2 have been parameterized by lower length scale simulations. The code can be used to simulate defect evolution in a prescribed UO2 microstructure as well as to solve defect clustering problems that control effective diffusivities under both thermal and irradiation conditions. It I possible to extend the current UO2 model to other fuel types such as accident tolerant fuels based on the U3Si2 compound. This would obviously require a new set of material properties describing the behavior of defects in U3Si2 rather than UO2. The framework is however designed to be generic.

  13. Theoretical Description of the Fission Process

    SciTech Connect

    Witold Nazarewicz

    2009-10-25

    Advanced theoretical methods and high-performance computers may finally unlock the secrets of nuclear fission, a fundamental nuclear decay that is of great relevance to society. In this work, we studied the phenomenon of spontaneous fission using the symmetry-unrestricted nuclear density functional theory (DFT). Our results show that many observed properties of fissioning nuclei can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. From the calculated collective potential and collective mass, we estimated spontaneous fission half-lives, and good agreement with experimental data was found. We also predicted a new phenomenon of trimodal spontaneous fission for some transfermium isotopes. Our calculations demonstrate that fission barriers of excited superheavy nuclei vary rapidly with particle number, pointing to the importance of shell effects even at large excitation energies. The results are consistent with recent experiments where superheavy elements were created by bombarding an actinide target with 48-calcium; yet even at high excitation energies, sizable fission barriers remained. Not only does this reveal clues about the conditions for creating new elements, it also provides a wider context for understanding other types of fission. Understanding of the fission process is crucial for many areas of science and technology. Fission governs existence of many transuranium elements, including the predicted long-lived superheavy species. In nuclear astrophysics, fission influences the formation of heavy elements on the final stages of the r-process in a very high neutron density environment. Fission applications are numerous. Improved understanding of the fission process will enable scientists to enhance the safety and reliability of the nation’s nuclear stockpile and nuclear reactors. The deployment of a fleet of safe and efficient advanced reactors, which will also minimize radiotoxic

  14. The phebus fission product project

    NASA Astrophysics Data System (ADS)

    von der Hardt, P.; Tattegrain, A.

    1992-06-01

    A new facility is being built at the Phebus test reactor in Cadarache, France, for investigations into phenomena of fuel damage and fission product (FP) release under severe power reactor accident conditions, as part of a large international research program. Phebus FP simulates core, cooling system and containment of an accidented reactor by appropriate scaled-down experimental components. The test fuel, with 25 to 30 GWd/t burnup, is re-irradiated in situ and then overheated up to UO 2 melting. Fission products and other aerosols are swept through the primary pipework into the containment vessel, by hot steam and hydrogen. Experimental instrumentation and posttest analyses will enable the following main phenomena to be studied: structural material and fuel dislocation, final fuel state; release, chemical form and transport/depletion of fission products in the facility, particularly aerosol physics, including nonfission product material and iodine chemistry in terms of volatile species formation through radiolysis, reactions with organic material, aerosol-vapor reactions, etc. Design and development of equipment and experimental procedures are supported by modeling and code calculations with the scope of predicting the experimental sequence, on one hand, and to prepare code validation through the results, on the other hand. More than 25 organisation from Europe and overseas, collaborate in the scientific and technological development of the Phebus FP program. The first in-pile test is planned for spring 1993, and five subsequent experiments are scheduled to follow in yearly intervals. This paper describes facility and support activities, and highlights a number of nuclear materials aspects involved.

  15. Factors controlling metal fuel lifetime

    SciTech Connect

    Porter, D.L.; Hofman, G.L.; Seidel, B.R.; Walters, L.C.

    1986-01-01

    The reliability of metal fuel elements is determined by a fuel burnup at which a statistically predicted number of fuel breaches would occur, the number of breaches determined by the amount of free fission gas which a particular reactor design can tolerate. The reliability is therefore measured using experimentally determined breach statistics, or by modelling fuel element behavior and those factors which contribute to cladding breach. The factors are fuel/cladding mechanical and chemical interactions, fission gas pressure, fuel phase transformations involving volume changes, and fission product effects on cladding integrity. Experimental data for EBR-II fuel elements has shown that the primary, and perhaps the only significant factor affecting metal fuel reliability, is the pressure-induced stresses caused by fission gas release. Other metal fuel/cladding systems may perform similarly.

  16. Fission products behaviour during a power transient: Their inventory in an intragranular bubble

    NASA Astrophysics Data System (ADS)

    Desgranges, L.; Blay, Th.; Lamontagne, J.; Roure, I.; Bienvenu, Ph.

    2017-09-01

    The behaviour of fission products is a key issue during Anticipated Operational Occurrences (AOOs) or Condition II transients or accidental sequence for nuclear fuel. Here we characterized how fission products behaved inside chromium doped UO2 pellet during a power ramp. At the pellet centre fission products have left the UO2 lattice and can be found in bubbles. The composition of the bubbles was determined using an original experimental methodology. The existence of separated precipitates made of metallic fission products for the one, and volatile fission products for the other, was evidenced. This result is discussed with regards to the behaviour of fission products during a power ramp.

  17. Effect of the β decay of metallic fission products on the chemical and phase compositions of the uranium-plutonium nitride nuclear fuel irradiated by fast neutrons

    NASA Astrophysics Data System (ADS)

    Bondarenko, G. G.; Androsov, A. V.; Bulatov, G. S.; Gedgovd, K. N.; Lyubimov, D. Yu.; Yakunkin, M. M.

    2016-09-01

    Thermodynamic analysis of the chemical and phase compositions of uranium-plutonium nitride (U0.8Pu0.2)N0.995 irradiated by fast neutrons to a burn-up fraction of 14% shows that a structure, which consists of a solid solution based on uranium and plutonium nitrides and containing some elements (americium, neptunium, zirconium, yttrium, lanthanides), individual condensed phases (U2N3, CeRu2, Ba3N2, CsI, Sr3N2, LaSe), metallic molybdenum and technetium, and U(Ru, Rh, Pd)3 intermetallics, forms due to the accumulation of metallic fission products. The contents and compositions of these phases are calculated. The change in the chemical and phase compositions of the irradiated uranium-plutonium nitride during the β decay of metallic radioactive fission products is studied. The kinetics of the transformations of 95Nb41N, 143Pr59N, 151Sm62N, and 147NdN into 95Mo42 + Ns.s., 143Nd60N, 151Eu63N, and 147SmN, respectively, is calculated.

  18. Liquid uranium alloy-helium fission reactor

    DOEpatents

    Minkov, V.

    1984-06-13

    This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

  19. NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1962-08-14

    A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

  20. Sensitivity analysis of the fission gas behavior model in BISON.

    SciTech Connect

    Swiler, Laura Painton; Pastore, Giovanni; Perez, Danielle; Williamson, Richard

    2013-05-01

    This report summarizes the result of a NEAMS project focused on sensitivity analysis of a new model for the fission gas behavior (release and swelling) in the BISON fuel performance code of Idaho National Laboratory. Using the new model in BISON, the sensitivity of the calculated fission gas release and swelling to the involved parameters and the associated uncertainties is investigated. The study results in a quantitative assessment of the role of intrinsic uncertainties in the analysis of fission gas behavior in nuclear fuel.

  1. Analysis of a Nuclear Accident: Fission and Activation Product Releases from the Fukushima Daiichi Nuclear Facility as Remote Indicators of Source Identification, Extent of Release, and State of Damaged Spent Nuclear Fuel

    SciTech Connect

    Schwantes, Jon M.; Orton, Christopher R.; Clark, Richard A.

    2012-09-10

    Measurements of several radionuclides within environmental samples taken from the Fukushima Daiichi nuclear facility and reported on the Tokyo Electric Power Company website following the recent tsunami-initiated catastrophe were evaluated for the purpose of identifying the source term, reconstructing the release mechanisms, and estimating the extent of the release. 136Cs/137Cs and 134Cs/137Cs ratios identified Units 1-3 as the major source of radioactive contamination to the surface soil close to the facility. A trend was observed between the fraction of the total core inventory released for a number of fission product isotopes and their corresponding Gibbs Free Energy of formation for the primary oxide form of the isotope, suggesting that release was dictated primarily by chemical volatility driven by temperature and reduction potential within the primary containment vessels of the vented reactors. The absence of any major fractionation beyond volatilization suggested all coolant had evaporated by the time of venting. High estimates for the fraction of the total inventory released of more volatile species (Te, Cs, I) indicated the damage to fuel bundles was likely extensive, minimizing any potential containment due to physical migration of these species through the fuel matrix and across the cladding wall. 238Pu/239,240Pu ratios close-in and at 30 km from the facility indicated that the damaged reactors were the major contributor of Pu to surface soil at the source but that this contribution likely decreased rapidly with distance from the facility. The fraction of the total Pu inventory released to the environment from venting units 1 and 3 was estimated to be ~0.003% based upon Pu/Cs isotope ratios relative to the within-reactor modeled inventory prior to venting and was consistent with an independent model evaluation that considered chemical volatility based upon measured fission product release trends. Significant volatile radionuclides within the spent fuel

  2. FUEL ELEMENT CONSTRUCTION

    DOEpatents

    Zumwalt, L.R.

    1961-08-01

    Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

  3. SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS FROM NEUTRON- BOMBARDED URANIUM

    DOEpatents

    Martin, A.E.; Johnson, I.; Burris, L. Jr.; Winsch, I.O.; Feder, H.M.

    1962-11-13

    A process is given for removing plutonium and/or fission products from uranium fuel. The fuel is dissolved in molten zinc--magnesium (10 to 18% Mg) alloy, more magnesium is added to obtain eutectic composition whereby uranium precipitates, and the uranium are separated from the Plutoniumand fission-product- containing eutectic. (AEC)

  4. Fission Reaction Event Yield Algorithm

    SciTech Connect

    Hagmann, Christian; Verbeke, Jerome; Vogt, Ramona; Roundrup, Jorgen

    2016-05-31

    FREYA (Fission Reaction Event Yield Algorithm) is a code that simulated the decay of a fissionable nucleus at specified excitation energy. In its present form, FREYA models spontaneous fission and neutron-induced fission up to 20 MeV. It includes the possibility of neutron emission from the nuclear prior to its fussion (nth chance fission).

  5. Data summary report for fission product release test VI-4

    SciTech Connect

    Obsorne, M.F.; Lorenz, R.A.; Collins, J.L.; Travis, J.R.; Webster, C.S.; Nakamura, T. )

    1991-01-01

    This was the fourth in a series of high-temperature fission product release tests in a vertical test apparatus. The test specimen, a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium, had been irradiated to a burnup of 47 MWd/kg. In simulation of a severe accident in a light-water reactor, it was heated in hydrogen in a hot cell-mounted test apparatus to a maximum test temperature of 2400 K for a period of 20 min. The released fission products were collected on components designed to facilitate sampling and analysis. On-line radioactivity measurements and posttest inspection revealed that the fuel had partially collapsed at about the time the cladding melted. Based on fission product inventories measured in the fuel or calculated by ORIGEN2, analyses of test components showed total releases from the fuel of 85% for {sup 85}Kr, <1% for {sup 106}Ru, 3.9% for {sup 125}Sb, 96% for both {sup 134}Cs and {sup 137}Cs, and 13% for {sup 154}Eu. Large fractions of the released fission products (up to 96% of the {sup 154}Eu) were retained in the furnace. Small release fractions for several other fission products -- Rb, Br, Sr, Te, I, and Ba -- were detected also. In addition, very small amounts of fuel material -- uranium and plutonium -- were released. Total mass release from the furnace to the collection system, which included fission products, fuel material, and structural materials, was 0.40g, with 40% of this material being deposited as vapor and 60% of it being collected as aerosols. The results from this test were compared with previous tests in this series and with an in-pile test at similar conditions at Sandia National Laboratories. There was no indication that the mode of heating (fission heat vs radiant heat) significantly affected fission product release. 24 refs., 25 figs., 14 tabs.

  6. Fission Product Release from SLOWPOKE-2 Reactors

    NASA Astrophysics Data System (ADS)

    Harnden, Anne M. C.

    Increasing radiation fields at several SLOWPOKE -2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace above the reactor were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements.

  7. Vented nuclear fuel element

    DOEpatents

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

    A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

  8. Fission Xenon on Mars

    NASA Technical Reports Server (NTRS)

    Mathew, K. J.; Marti, K.; Marty, B.

    2002-01-01

    Fission Xe components due to Pu-244 decay in the early history of Mars have been identified in nakhlites; as in the case of ALH84001 and Chassigny the fission gas was assimilated into indigenous solar-type Xe. Additional information is contained in the original extended abstract.

  9. Geometry of membrane fission.

    PubMed

    Frolov, Vadim A; Escalada, Artur; Akimov, Sergey A; Shnyrova, Anna V

    2015-01-01

    Cellular membranes define the functional geometry of intracellular space. Formation of new membrane compartments and maintenance of complex organelles require division and disconnection of cellular membranes, a process termed membrane fission. Peripheral membrane proteins generally control membrane remodeling during fission. Local membrane stresses, reflecting molecular geometry of membrane-interacting parts of these proteins, sum up to produce the key membrane geometries of fission: the saddle-shaped neck and hour-glass hemifission intermediate. Here, we review the fundamental principles behind the translation of molecular geometry into membrane shape and topology during fission. We emphasize the central role the membrane insertion of specialized protein domains plays in orchestrating fission in vitro and in cells. We further compare individual to synergistic action of the membrane insertion during fission mediated by individual protein species, proteins complexes or membrane domains. Finally, we describe how local geometry of fission intermediates defines the functional design of the protein complexes catalyzing fission of cellular membranes. Copyright © 2014 Elsevier Ireland Ltd. All rights reserved.

  10. The Fission Barrier Landscape

    SciTech Connect

    Phair, L.; Moretto, L. G.

    2008-04-17

    Fission excitation functions have been measured for a chain of neighboring compound nuclei from {sup 207}Po to {sup 212}Po. We present a new analysis which provides a determination of the fission barriers and ground state shell effects with nearly spectroscopic accuracy. The accuracy achieved in this analysis may lead to a future detailed exploration of the saddle mass surface and its spectroscopy.

  11. Biomodal spontaneous fission

    SciTech Connect

    Hulet, E.K. )

    1989-09-26

    Investigations of mass and kinetic-energy distributions from spontaneous fission have been extended in recent years to an isotope of element 104 and, for half-lives, to an isotope of element 108. The results have been surprising in that spontaneous fission half-lives have turned out to be much longer than expected and mass and kinetic- energy distributions were found to abruptly shift away from those of the lighter actinides, showing two modes of fission. These new developments have caused a re-evaluation of our understanding of the fission process, bringing an even deeper appreciation of the role played by nuclear shell effects upon spontaneous fission properties. 16 refs., 10 figs.

  12. Fission Measurements with Dance

    NASA Astrophysics Data System (ADS)

    Jandel, M.; Bredeweg, T. A.; Fowler, M. M.; Bond, E. M.; Chadwick, M. B.; Clement, R. R.; Couture, A.; O'Donnell, J. M.; Haight, R. C.; Keksis, A. L.; Reifarth, R.; Rundberg, R. S.; Ullmann, J. L.; Vieira, D. J.; Wilhelmy, J. B.; Wouters, J. M.; Agvaanluvsan, U.; Dashdorj, D.; Macri, R. A.; Parker, W. E.; Wilk, P. A.; Wu, C. Y.; Becker, J. A.; Angell, C. T.; Tonchev, A. P.; Baker, J. D.

    2008-08-01

    Neutron capture cross section measurements on actinides are complicated by the presence of neutron-induced fission. An efficient fission tagging detector used in coincidence with the Detector for Advanced Neutron Capture Experiments (DANCE) provides a powerful tool in undertaking simultaneous measurements of (n,γ) and (n,f) cross sections. Preliminary results on 235U(n,γ) and (n,f) and 242mAm(n,f) cross sections measured with DANCE and a custom fission-tagging parallel plate avalanche counter (PPAC) are presented. Additional measurements of γ-ray cluster multiplicity distributions for neutron-induced fission of 235U and 242mAm and spontaneous fission of 252Cf are shown, as well as γ-ray energy and average γ-ray energy distributions.

  13. Analysis of fission product release behavior during the TMI-2 accident

    SciTech Connect

    Petti, D. A.; Adams, J. P.; Anderson, J. L.; Hobbins, R. R.

    1987-01-01

    An analysis of fission product release during the Three Mile Island Unit 2 (TMI-2) accident has been initiated to provide an understanding of fission product behavior that is consistent with both the best estimate accident scenario and fission product results from the ongoing sample acquisition and examination efforts. ''First principles'' fission product release models are used to describe release from intact, disrupted, and molten fuel. Conclusions relating to fission product release, transport, and chemical form are drawn. 35 refs., 12 figs., 7 tabs.

  14. Multiscale development of a fission gas thermal conductivity model: Coupling atomic, meso and continuum level simulations

    NASA Astrophysics Data System (ADS)

    Tonks, Michael R.; Millett, Paul C.; Nerikar, Pankaj; Du, Shiyu; Andersson, David; Stanek, Christopher R.; Gaston, Derek; Andrs, David; Williamson, Richard

    2013-09-01

    Fission gas production and evolution significantly impact the fuel performance, causing swelling, a reduction in the thermal conductivity and fission gas release. However, typical empirical models of fuel properties treat each of these effects separately and uncoupled. Here, we couple a fission gas release model to a model of the impact of fission gas on the fuel thermal conductivity. To quantify the specific impact of grain boundary (GB) bubbles on the thermal conductivity, we use atomistic and mesoscale simulations. Atomistic molecular dynamic simulations were employed to determine the GB thermal resistance. These values were then used in mesoscale heat conduction simulations to develop a mechanistic expression for the effective GB thermal resistance of a GB containing gas bubbles, as a function of the percentage of the GB covered by fission gas. The coupled fission gas release and thermal conductivity model was implemented in Idaho National Laboratory's BISON fuel performance code to model the behavior of a 10-pellet LWR fuel rodlet, showing how the fission gas impacts the UO2 thermal conductivity. Furthermore, additional BISON simulations were conducted to demonstrate the impact of average grain size on both the fuel thermal conductivity and the fission gas release.

  15. Nuclear reactor fuel element

    DOEpatents

    Johnson, Carl E.; Crouthamel, Carl E.

    1980-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

  16. FUEL ELEMENT CONSTRUCTION

    DOEpatents

    Simnad, M.T.

    1961-08-15

    A method of preventing diffusible and volatile fission products from diffusing through a fuel element container and contaminating reactor coolant is described. More specifically, relatively volatile and diffusible fission products either are adsorbed by or react with magnesium fluoride or difluoride to form stable, less volatile, less diffusible forms. The magnesium fluoride or difluoride is disposed anywhere inwardly from the outer surface of the fuel element container in order to be contacted by the fission products before they reach and contaminate the reactor coolant. (AEC)

  17. Fission gas in thoria

    NASA Astrophysics Data System (ADS)

    Kuganathan, Navaratnarajah; Ghosh, Partha S.; Galvin, Conor O. T.; Arya, Ashok K.; Dutta, Bijon K.; Dey, Gautam K.; Grimes, Robin W.

    2017-03-01

    The fission gases Xe and Kr, formed during normal reactor operation, are known to degrade fuel performance, particularly at high burn-up. Using first-principles density functional theory together with a dispersion correction (DFT + D), in ThO2 we calculate the energetics of neutral and charged point defects, the di-vacancy (DV), different neutral tri-vacancies (NTV), the charged tetravacancy (CTV) defect cluster geometries and their interaction with Xe and Kr. The most favourable incorporation point defect site for Xe or Kr in defective ThO2 is the fully charged thorium vacancy. The lowest energy NTV in larger supercells of ThO2 is NTV3, however, a single Xe atom is most stable when accommodated within a NTV1. The di-vacancy (DV) is a significantly less favoured incorporation site than the NTV1 but the CTV offers about the same incorporation energy. Incorporation of a second gas atom in a NTV is a high energy process and more unfavourable than accommodation within an existing Th vacancy. The bi-NTV (BNTV) cluster geometry studied will accommodate one or two gas atoms with low incorporation energies but the addition of a third gas atom incurs a high energy penalty. The tri-NTV cluster (TNTV) forms a larger space which accommodates three gas atoms but again there is a penalty to accommodate a fourth gas atom. By considering the energy to form the defect sites, solution energies were generated showing that in ThO2-x the most favourable solution equilibrium site is the NTV1 while in ThO2 it is the DV.

  18. Fissioning uranium plasmas and nuclear-pumped lasers

    NASA Technical Reports Server (NTRS)

    Schneider, R. T.; Thom, K.

    1975-01-01

    Current research into uranium plasmas, gaseous-core (cavity) reactors, and nuclear-pumped lasers is discussed. Basic properties of fissioning uranium plasmas are summarized together with potential space and terrestrial applications of gaseous-core reactors and nuclear-pumped lasers. Conditions for criticality of a uranium plasma are outlined, and it is shown that the nonequilibrium state and the optical thinness of a fissioning plasma can be exploited for the direct conversion of fission fragment energy into coherent light (i.e., for nuclear-pumped lasers). Successful demonstrations of nuclear-pumped lasers are described together with gaseous-fuel reactor experiments using uranium hexafluoride.

  19. Bimodal fission of Hs*

    NASA Astrophysics Data System (ADS)

    Itkis, I. M.; Itkis, M. G.; Knyazheva, G. N.; Kozulin, E. M.; Loktev, T. A.; Novikov, K. V.; Hanappe, F.; Vardaci, E.

    2014-05-01

    Mass and energy distributions of fission fragments obtained in the reactions 22Ne + 249Cf, 26Mg + 248Cm, and 22Ne + 238U have been measured. A special attention will be paid on the properties of mass-energy distribution of fission fragments obtained in the reaction 26Mg + 248Cm at an excitation energy of 35 MeV. At this energy shell effects should become more effective in fission, the TKE distribution of symmetric fragments obtained in the reaction 26Mg + 248Cm differs strongly from a Gaussian shape. Besides a low-energy component, a high-energy component, not foreseen in the LDM, arises. This is attributed to the fact that both fission fragments are close to the spherical neutron shell N = 82. It means that for the compound nucleus 274Hs*, formed in the reaction 26Mg + 248Cm, the phenomenon of bimodal fission was observed for the first time. For the compound nucleus 260No* formed in the reaction 22Ne + 238U at the initial excitation energy of 41 MeV the bimodal fission as well as superasymmetric fission were observed.

  20. Nuclear fission of Fm isotopes

    SciTech Connect

    Asano, T.; Wada, T.; Ohta, M.; Chiba, S.

    2010-06-01

    Multi-modal fission has been systematically investigated for the series of isotopes of Fm and Cf. The multi-dimensional Langevin-type stochastic differential equation is used for the dynamical calculation. The primary fission mode changes from mass-asymmetric fission to mass-symmetric fission with the increase of neutron numbers for both Fm and Cf cases.

  1. FUEL ELEMENT

    DOEpatents

    Bean, R.W.

    1963-11-19

    A ceramic fuel element for a nuclear reactor that has improved structural stability as well as improved cooling and fission product retention characteristics is presented. The fuel element includes a plurality of stacked hollow ceramic moderator blocks arranged along a tubular raetallic shroud that encloses a series of axially apertured moderator cylinders spaced inwardly of the shroud. A plurality of ceramic nuclear fuel rods are arranged in the annular space between the shroud and cylinders of moderator and appropriate support means and means for directing gas coolant through the annular space are also provided. (AEC)

  2. Steady-state fuel behavior modeling of nitride fuels in FRAPCON-EP

    NASA Astrophysics Data System (ADS)

    Feng, Bo; Karahan, Aydın; Kazimi, Mujid S.

    2012-08-01

    Fuel material properties and mechanistic fission gas models in FRAPCON-EP were updated to model the steady-state behavior of high-porosity nitride fuel operating at temperatures below half of the melting point. The fuel thermal conductivity and fuel thermal expansion models were updated with correlations for UN and (U,Pu)N fuels. Hot-pressing of the as-fabricated porosity was modeled as a function of the hydrostatic pressure and creep rate. The solid fission product swelling was assumed to increase linearly with burnup. Fission gas swelling constitutive models were updated to appropriately capture the intragranular gas bubble evolution in nitride fuel. Intergranular gas swelling was neglected due to the assumed high porosity of the fuel. The fission gas release behavior was modeled by fitting the fission gas diffusion coefficient in UN to FRAPCON's default fission gas release model. This fitted gas diffusion coefficient reflects the effects of porosity, burnup, operating temperature, fission rate, and bubble sink strength. Fission gas release and fuel swelling benchmarks against irradiation data were performed. The updated code was applied to UN fuel in typical PWR geometry and operating conditions, with an extended cycle length of 24 months. The results show that swelling of the nitride fuel up to 60 MWd/kg burnup did not lead to excessive straining of the cladding. Furthermore, this study showed that a porous (>15% porosity) nitride fuel pellet could achieve a much higher margin to failure from the cladding collapse and grid-to-rod fretting.

  3. Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu, 235U, 238U

    NASA Astrophysics Data System (ADS)

    Selby, H. D.; Mac Innes, M. R.; Barr, D. W.; Keksis, A. L.; Meade, R. A.; Burns, C. J.; Chadwick, M. B.; Wallstrom, T. C.

    2010-12-01

    We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for 99Mo, 95Zr, 137Cs, 140Ba, 141,143Ce, and 147Nd. Modest incident-energy dependence exists for the 147Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by ˜5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except for 99Mo

  4. Development Progress in Phase 1 Fission Propulsion Systems

    NASA Technical Reports Server (NTRS)

    Houts, Michael G.; VanDyke, Melissa; Godfroy, Tom; Martin, Jim; Dickens, Ricky; Pedersen, Kevin; Poston, David; Reid, Bob; Lipinski, Ron; Wright, Steve; hide

    2000-01-01

    Phase 1 fission propulsion systems are those fission propulsion systems that are highly testable and require no development of nuclear fuels or materials. The systems can be developed without new or significantly modified facilities, have adequate performance for numerous missions of interest, and demonstrate technologies and programmatics that are traceable to Phase 2 and Phase 3 systems. Phase 1 fission propulsion systems focus on safety, cost and schedule. Phase 1 flight units can be tested at full thrust using resistance heaters to simulate heat from fission. The development and use of Phase 1 systems will help enable Phase 2 or Phase 3 fission propulsion systems capable of giving rapid, affordable access to any point in the solar system. A Phase 1 fission propulsion system under development at the Marshall Space Flight Center (MSFC) in collaboration with individuals from Department of Energy Laboratories and industry is the Safe Affordable Fission Engine (SAFE). The propellant energy source of a 30 kW SAFE unit (SAFE-30) is being fabricated, and will begin testing at MSFC in FY00. The conceptual design of a 300 kW SAFE unit (SAFE-300)is nearing completion. Experiments have been performed on both SAFE-30 and SAFE-300 components. Module tests have confirmed the performance potential of the SAFE series of propulsion systems. This paper will report on the development status of the Phase 1 SAFE fission propulsion system.

  5. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-11-21

    A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

  6. Fission Surface Power Technology Development Update

    NASA Technical Reports Server (NTRS)

    Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

    2011-01-01

    Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power

  7. Fission Systems for Mars Exploration

    NASA Technical Reports Server (NTRS)

    Houts, Michael G.; Kim, T.; Dorney, D. J.; Swint, Marion Shayne

    2012-01-01

    Fission systems are used extensively on earth, and 34 such systems have flown in space. The energy density of fission is over 10 million times that of chemical reactions, giving fission the potential to eliminate energy density constraints for many space missions. Potential safety and operational concerns with fission systems are well understood, and strategies exist for affordably developing such systems. By enabling a power-rich environment and highly efficient propulsion, fission systems could enable affordable, sustainable exploration of Mars.

  8. Fission product release mechanisms and pathways

    SciTech Connect

    Malinauskas, A.P.

    1981-01-01

    It is axiomatic that the severity of a nuclear reactor accident is determined by the extent of radioactivity escape which results. The main focus of site safety analyses is thus on fission product release and transport. Of all the processes involved, fission product escape from the fuel-cladding region into the primary coolant circuit is perhaps the most simple to describe; even so, it is an extremely complex function of the time/temperature history of the fuel-cladding system during an accident, since many mechanisms for release are involved. Depending upon the particular fission product species, these release mechanisms range from simple gaseous expansion processes at low temperatures to evaporation-condensation processes (aerosol formation) over molten fuel. Because of these complexities, it is convenient to subdivide the time/temperature sequence of an accident into more or less discrete phases over which specific release mechanisms dominate. Four such phases are the periods of (1) gap release, (2) meltdown release, (3) vaporization, and (4) oxidation release. This approach simplifies the problem considerably, although some loss of uniformity results. The methodology applies to BWR and PWR reactors with appropriate adaptations.

  9. True ternary fission

    NASA Astrophysics Data System (ADS)

    Vijayaraghavan, K. R.; Balasubramaniam, M.; von Oertzen, W.

    2015-04-01

    The study of the ternary fission of nuclei has received new interest recently. It is of general interest for nuclear dynamics, although the process is very rare. In the present work, we discuss the possibilities of true ternary fission (fragment masses A >30 ) in 252Cf for different mass splits. These mass splits are strongly favored in a collinear geometry. Based on the three cluster model (TCM), it is shown that the true ternary fission into fragments with almost equal masses is one of the possible fission modes in 252Cf . For general decays it is shown that the formation of the lightest fragment at the center has the highest probability. Further the formation of tin isotopes and/or other closed shell fragments are favored. For the decay products the presence of closed shell nuclei among the three fragments enhances the decay probabilities.

  10. Fission in a Plasma

    SciTech Connect

    Younes, W.

    2016-10-26

    A three-year theory project was undertaken to study the fission process in extreme astrophysical environments, such as the crust of neutron stars. In the first part of the project, the effect of electron screening on the fission process was explored using a microscopic approach. For the first time, these calculations were carried out to the breaking point of the nucleus. In the second part of the project, the population of the fissioning nucleus was calculated within the same microscopic framework. These types of calculations are extremely computer-intensive and have seldom been applied to heavy deformed nuclei, such as fissioning actinides. The results, tools and methodologies produced in this work will be of interest to both the basic-science and nuclear-data communities.

  11. Microscopic Theory of Fission

    SciTech Connect

    Younes, W.; Gogny, D.

    2008-04-17

    In recent years, the microscopic method has been applied to the notoriously difficult problem of nuclear fission with unprecedented success. In this paper, we discuss some of the achievements and promise of the microscopic method, as embodied in the Hartree-Fock method using the Gogny finite-range effective interaction, and beyond-mean-field extensions to the theory. The nascent program to describe induced fission observables using this approach at the Lawrence Livermore National Laboratory is presented.

  12. Background and Derivation of ANS-5.4 Standard Fission Product Release Model

    SciTech Connect

    Beyer, Carl E.; Turnbull, Andrew J.

    2010-01-29

    This background report describes the technical basis for the newly proposed American Nuclear Society (ANS) 5.4 standard, Methods for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuels. The proposed ANS 5.4 standard provides a methodology for determining the radioactive fission product releases from the fuel for use in assessing radiological consequences of postulated accidents that do not involve abrupt power transients. When coupled with isotopic yields, this method establishes the 'gap activity,' which is the inventory of volatile fission products that are released from the fuel rod if the cladding are breached.

  13. Singlet exciton fission photovoltaics.

    PubMed

    Lee, Jiye; Jadhav, Priya; Reusswig, Philip D; Yost, Shane R; Thompson, Nicholas J; Congreve, Daniel N; Hontz, Eric; Van Voorhis, Troy; Baldo, Marc A

    2013-06-18

    Singlet exciton fission, a process that generates two excitons from a single photon, is perhaps the most efficient of the various multiexciton-generation processes studied to date, offering the potential to increase the efficiency of solar devices. But its unique characteristic, splitting a photogenerated singlet exciton into two dark triplet states, means that the empty absorption region between the singlet and triplet excitons must be filled by adding another material that captures low-energy photons. This has required the development of specialized device architectures. In this Account, we review work to develop devices that harness the theoretical benefits of singlet exciton fission. First, we discuss singlet fission in the archetypal material, pentacene. Pentacene-based photovoltaic devices typically show high external and internal quantum efficiencies. They have enabled researchers to characterize fission, including yield and the impact of competing loss processes, within functional devices. We review in situ probes of singlet fission that modulate the photocurrent using a magnetic field. We also summarize studies of the dissociation of triplet excitons into charge at the pentacene-buckyball (C60) donor-acceptor interface. Multiple independent measurements confirm that pentacene triplet excitons can dissociate at the C60 interface despite their relatively low energy. Because triplet excitons produced by singlet fission each have no more than half the energy of the original photoexcitation, they limit the potential open circuit voltage within a solar cell. Thus, if singlet fission is to increase the overall efficiency of a solar cell and not just double the photocurrent at the cost of halving the voltage, it is necessary to also harvest photons in the absorption gap between the singlet and triplet energies of the singlet fission material. We review two device architectures that attempt this using long-wavelength materials: a three-layer structure that uses

  14. Migration of fission products in UO{sub 2}. Final report

    SciTech Connect

    Prussin, S.G.; Olander, D.R.

    1995-12-01

    Results of an experimental and calculational effort to examine the fundamental mechanisms of fission product migration in and release from polycrystalline uranium dioxide are reported. The experiments were designed to provide diffusion parameters for the representative fission products tellurium, iodine, xenon, molybdenum and ruthenium under both reducing and oxidizing conditions. The calculational effort applied a new model of fission product release from reactor fuel that incorporates grain growth as well as grain boundary and lattice diffusion.

  15. Improved Fission Neutron Data Base for Active Interrogation of Actinides

    SciTech Connect

    Pozzi, Sara; Czirr, J. Bart; Haight, Robert; Kovash, Michael; Tsvetkov, Pavel

    2013-11-06

    This project will develop an innovative neutron detection system for active interrogation measurements. Many active interrogation methods to detect fissionable material are based on the detection of neutrons from fission induced by fast neutrons or high-energy gamma rays. The energy spectrum of the fission neutrons provides data to identify the fissionable isotopes and materials such as shielding between the fissionable material and the detector. The proposed path for the project is as follows. First, the team will develop new neutron detection systems and algorithms by Monte Carlo simulations and bench-top experiments. Next, They will characterize and calibrate detection systems both with monoenergetic and white neutron sources. Finally, high-fidelity measurements of neutron emission from fissions induced by fast neutrons will be performed. Several existing fission chambers containing U-235, Pu-239, U-238, or Th-232 will be used to measure the neutron-induced fission neutron emission spectra. The challenge for making confident measurements is the detection of neutrons in the energy ranges of 0.01 – 1 MeV and above 8 MeV, regions where the basic data on the neutron energy spectrum emitted from fission is least well known. In addition, improvements in the specificity of neutron detectors are required throughout the complete energy range: they must be able to clearly distinguish neutrons from other radiations, in particular gamma rays and cosmic rays. The team believes that all of these challenges can be addressed successfully with emerging technologies under development by this collaboration. In particular, the collaboration will address the area of fission neutron emission spectra for isotopes of interest in the advanced fuel cycle initiative (AFCI).

  16. Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

    SciTech Connect

    Gauld, I.C.

    2005-08-12

    U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd.

  17. Control of a laser inertial confinement fusion-fission power plant

    DOEpatents

    Moses, Edward I.; Latkowski, Jeffery F.; Kramer, Kevin J.

    2015-10-27

    A laser inertial-confinement fusion-fission energy power plant is described. The fusion-fission hybrid system uses inertial confinement fusion to produce neutrons from a fusion reaction of deuterium and tritium. The fusion neutrons drive a sub-critical blanket of fissile or fertile fuel. A coolant circulated through the fuel extracts heat from the fuel that is used to generate electricity. The inertial confinement fusion reaction can be implemented using central hot spot or fast ignition fusion, and direct or indirect drive. The fusion neutrons result in ultra-deep burn-up of the fuel in the fission blanket, thus enabling the burning of nuclear waste. Fuels include depleted uranium, natural uranium, enriched uranium, spent nuclear fuel, thorium, and weapons grade plutonium. LIFE engines can meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the highly undesirable stockpiles of depleted uranium, spent nuclear fuel and excess weapons materials.

  18. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1963-06-11

    A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

  19. FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Carney, K.G. Jr.

    1959-07-14

    A nuclear fuel element comprising a large number og wafers of fissionable material and a protective jacket having compartments holding these wafers is described. The compartments of the jacket aid the removal of heat from the wafers, keep the wafers or fragments thereof from migrating in the jacket, and permit the escape of gaseous fission products.

  20. Modernizing the Fission Basis

    NASA Astrophysics Data System (ADS)

    Tonchev, Anton; Henderson, Roger; Schunck, Nicolas; Sroyer, Mark; Vogt, Ramona

    2016-09-01

    In 1939, Niels Bohr and John Wheeler formulated a theory of neutron-induced nuclear fission based on the hypothesis of the compound nucleus. Their theory, the so-called ``Bohr hypothesis,'' is still at the heart of every theoretical fission model today and states that the decay of a compound nucleus for a given excitation energy, spin, and parity is independent of its formation. We propose the first experiment to validate to 1-2% absolute uncertainties the practical consequences of the Bohr hypothesis during induced nuclear fission. We will compare the fission product yields (FPYs) of the same 240Pu compound nucleus produced via two different reactions (i) n+239Pu and (ii) γ+240 Pu. These high-precision FPYs measurements will be extremely beneficial for our fundamental understanding of the nuclear fission process and nuclear reactions from first principles. This work was performed under the auspices of US DOE by LLNL under Contract DE-AC52-07NA27344. Funding was provided via the LDRD-ERD-069 project.

  1. Data summary report for fission product release test VI-3

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Collins, J.L.; Travis, J.R.; Webster, C.S.; Lee, H.K.; Nakamura, T.; Tong, Y.-C. )

    1990-06-01

    Test VI-3, the third in a series of high-temperature fission product release tests in the vertical test apparatus, was conducted in flowing steam. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium, which had been irradiated to a burnup of 42 MWd/kg. Using an induction furnace, it was heated under simulated light-water reactor (LWR) accident conditions to two test temperatures, 20 min at 2000 K and then 20 min at 2700 K, in a hot cell-mounted test apparatus. The released fission products were collected on components designed to facilitate sampling and analysis. Posttest inspection confirmed that the cladding had been completely oxidized during the test. Only minimal fragmentation of the fuel specimen was found, however, and very little melting or fuel-cladding interaction had occurred. Based on fission product inventories measured in the fuel or calculated by ORIGEN2, analyses of test components showed total releases from the fuel of 100% for {sup 85}Kr, 5% for {sup 106}Ru, 99% for {sup 125}Sb, and 99% for both {sup 134}Cs and {sup 137}Cs. A large fraction (27%) of the released {sup 125}Sb was retained in the furnace, but most of the released cesium (89%) escaped to the collection system. In addition, very small amounts of fuel material --- uranium and plutonium --- were released. Including fission products and fuel and structural materials, the total mass released from the furnace to the collection system was 3.17 g, 78% of which was collected on the filters. The results from this test were compared with previous tests in this series and with a commonly used model for fission product release. 25 refs., 22 figs., 14 tabs.

  2. Feasibility of an antiproton catalyzed fission fragment rocket

    SciTech Connect

    Hdinger, D.S.

    1992-03-01

    The purpose of this project was to investigate the feasibility of an antiproton catalyzed fission fragment rocket (FFR). The FFR is characterized by the extraction of fission fragments from the fissile fuel, and the utilization of their kinetic energy for thrust generation. A significant drawback to previous FFR designs was the requirement to maintain a critical nuclear pile as the fission fragment source. The author examined the possibility of replacing the critical pile with a sub-critical pile driven by antiprotons. Recent experiments have revealed that antiprotons stimulate highly energetic fissions in {sup 238}U, with a neutron multiplicity of 13.7 neutrons per fission. This interaction was used as a throttled neutron source. The pile consisted of layers of fissile coated fibers which are designed to allow fission fragments to escape them, where the fragments collide with a fluid. The heated fluid is then ejected from the rocket to provide thrust. The calculations performed indicate that each antiproton injected into the pile can stimulate 8 or more fissions while maintaining a neutron multiplication of less than 0.4. Based on the results seen, the engine design presented is inadequate. Limitations introduced by the reaction fluid far outweigh the simplicity-of-design gained. Despite this, the basic idea of using the antiproton-U interaction as a source of spacecraft propulsion warrants further study.

  3. Thermodynamics of fission products in UO2+-x

    SciTech Connect

    Nerikar, Pankaj V

    2009-01-01

    The stabilities of selected fission products - Xe, Cs, and Sr - are investigated as a function of non-stoichiometry x in UO{sub 2{+-}x}. In particular, density functional theory (OFT) is used to calculate the incorporation and solution energies of these fission products at the anion and cation vacancy sites, at the divacancy, and at the bound Schottky defect. In order to reproduce the correct insulating state of UO{sub 2}, the DFT calculations are performed using spin polarization and with the Hubbard U tenn. In general, higher charge defects are more soluble in the fuel matrix and the solubility of fission products increases as the hyperstoichiometry increases. The solubility of fission product oxides is also explored. CS{sub 2}O is observed as a second stable phase and SrO is found to be soluble in the UO{sub 2} matrix for all stoichiometries. These observations mirror experimentally observed phenomena.

  4. Fusion-fission hybrid studies in the United States

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Berwald, D.H.; Cheng, E.T.; Delene, J.G.; Jassby, D.L.

    1986-05-20

    Systems and conceptual design studies have been carried out on the following three hybrid types: (1) The fission-suppressed hybrid, which maximizes fissile material produced (Pu or /sup 233/U) per unit of total nuclear power by suppressing the fission process and multiplying neutrons by (n,2n) reactions in materials like beryllium. (2) The fast-fission hybrid, which maximizes fissile material produced per unit of fusion power by maximizing fission of /sup 238/U (Pu is produced) in which twice the fissile atoms per unit of fusion power (but only a third per unit of nuclear power) are made. (3) The power hybrid, which amplifies power in the blanket for power production but does not produce fuel to sell. All three types must sell electrical power to be economical.

  5. Fission Product Library and Resource

    SciTech Connect

    Burke, J. T.; Padgett, S.

    2016-09-29

    Fission product yields can be extracted from an irradiated sample by performing gamma ray spectroscopy on the whole sample post irradiation. There are several pitfalls to avoid when trying to determine a specific isotope's fission product yield.

  6. FUEL ELEMENTS FOR NEUTRONIC REACTORS

    DOEpatents

    Foote, F.G.; Jette, E.R.

    1963-05-01

    A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

  7. Process for treating fission waste

    DOEpatents

    Rohrmann, Charles A.; Wick, Oswald J.

    1983-01-01

    A method is described for the treatment of fission waste. A glass forming agent, a metal oxide, and a reducing agent are mixed with the fission waste and the mixture is heated. After melting, the mixture separates into a glass phase and a metal phase. The glass phase may be used to safely store the fission waste, while the metal phase contains noble metals recovered from the fission waste.

  8. ORNL fission product release tests VI-6

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Collins, J.L.; Lee, C.S.

    1991-01-01

    The ORNL fission product release tests investigate release and transport of the major fission products from high-burnup fuel under LWR accident conditions. The two most recent tests (VI-4 and VI-5) were conducted in hydrogen. In three previous tests in this series (VI-1, VI-2, and VI-3), which had been conducted in steam, the oxidized Zircaloy cladding remained largely intact and acted as a barrier to steam reaction with the UO{sub 2}. Test VI-6 was designed to insure significant oxidation of the UO{sub 2} fuel, which has been shown to enhance release of certain fission products, especially molybdenum and ruthenium. The BR3 fuel specimen used in test VI-6 will be heated in hydrogen to 2300 K; the Zircaloy cladding is expected to melt and runoff at {approximately}2150 K. Upon reaching the 2300 K test temperature, the test atmosphere will be changed to steam, and that temperature will be maintained for 60 min, with the three collection trains being operated for 2-, 18-, and 40-min periods. The releases of {sup 85}Kr and {sup 137}Cs will be monitored continuously throughout the test. Posttest analyses of the material collected on the three trains will provide results on the release and transport of Mo, Ru, Sb, Te, Ba, Ce, and Eu as a function of time at 2300 K. Continuous monitoring of the hydrogen produced during the steam atmosphere period at high temperature will provide a measure of the oxidation rate of the cladding and fuel. Following delays in approval of the safety documentation and in decontamination of the hot cell and test apparatus, test VI-6 will be conducted in late May.

  9. Fission modelling with FIFRELIN

    NASA Astrophysics Data System (ADS)

    Litaize, Olivier; Serot, Olivier; Berge, Léonie

    2015-12-01

    The nuclear fission process gives rise to the formation of fission fragments and emission of particles (n,γ , e-) . The particle emission from fragments can be prompt and delayed. We present here the methods used in the FIFRELIN code, which simulates the prompt component of the de-excitation process. The methods are based on phenomenological models associated with macroscopic and/or microscopic ingredients. Input data can be provided by experiment as well as by theory. The fission fragment de-excitation can be performed within Weisskopf (uncoupled neutron and gamma emission) or a Hauser-Feshbach (coupled neutron/gamma emission) statistical theory. We usually consider five free parameters that cannot be provided by theory or experiments in order to describe the initial distributions required by the code. In a first step this set of parameters is chosen to reproduce a very limited set of target observables. In a second step we can increase the statistics to predict all other fission observables such as prompt neutron, gamma and conversion electron spectra but also their distributions as a function of any kind of parameters such as, for instance, the neutron, gamma and electron number distributions, the average prompt neutron multiplicity as a function of fission fragment mass, charge or kinetic energy, and so on. Several results related to different fissioning systems are presented in this work. The goal in the next decade will be i) to replace some macroscopic ingredients or phenomenological models by microscopic calculations when available and reliable, ii) to be a support for experimentalists in the design of detection systems or in the prediction of necessary beam time or count rates with associated statistics when measuring fragments and emitted particle in coincidence iii) extend the model to be able to run a calculation when no experimental input data are available, iv) account for multiple chance fission and gamma emission before fission, v) account for the

  10. Fission Matrix Capability for MCNP Monte Carlo

    SciTech Connect

    Carney, Sean E.; Brown, Forrest B.; Kiedrowski, Brian C.; Martin, William R.

    2012-09-05

    In a Monte Carlo criticality calculation, before the tallying of quantities can begin, a converged fission source (the fundamental eigenvector of the fission kernel) is required. Tallies of interest may include powers, absorption rates, leakage rates, or the multiplication factor (the fundamental eigenvalue of the fission kernel, k{sub eff}). Just as in the power iteration method of linear algebra, if the dominance ratio (the ratio of the first and zeroth eigenvalues) is high, many iterations of neutron history simulations are required to isolate the fundamental mode of the problem. Optically large systems have large dominance ratios, and systems containing poor neutron communication between regions are also slow to converge. The fission matrix method, implemented into MCNP[1], addresses these problems. When Monte Carlo random walk from a source is executed, the fission kernel is stochastically applied to the source. Random numbers are used for: distances to collision, reaction types, scattering physics, fission reactions, etc. This method is used because the fission kernel is a complex, 7-dimensional operator that is not explicitly known. Deterministic methods use approximations/discretization in energy, space, and direction to the kernel. Consequently, they are faster. Monte Carlo directly simulates the physics, which necessitates the use of random sampling. Because of this statistical noise, common convergence acceleration methods used in deterministic methods do not work. In the fission matrix method, we are using the random walk information not only to build the next-iteration fission source, but also a spatially-averaged fission kernel. Just like in deterministic methods, this involves approximation and discretization. The approximation is the tallying of the spatially-discretized fission kernel with an incorrect fission source. We address this by making the spatial mesh fine enough that this error is negligible. As a consequence of discretization we get a

  11. Fusion-fission energy systems evaluation

    SciTech Connect

    Teofilo, V.L.; Aase, D.T.; Bickford, W.E.

    1980-01-01

    This report serves as the basis for comparing the fusion-fission (hybrid) energy system concept with other advanced technology fissile fuel breeding concepts evaluated in the Nonproliferation Alternative Systems Assessment Program (NASAP). As such, much of the information and data provided herein is in a form that meets the NASAP data requirements. Since the hybrid concept has not been studied as extensively as many of the other fission concepts being examined in NASAP, the provided data and information are sparse relative to these more developed concepts. Nevertheless, this report is intended to provide a perspective on hybrids and to summarize the findings of the rather limited analyses made to date on this concept.

  12. System Concepts for Affordable Fission Surface Power

    NASA Technical Reports Server (NTRS)

    Mason, Lee; Poston, David; Qualls, Louis

    2008-01-01

    This paper presents an overview of an affordable Fission Surface Power (FSP) system that could be used for NASA applications on the Moon and Mars. The proposed FSP system uses a low temperature, uranium dioxide-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The concept was determined by a 12 month NASA/DOE study that examined design options and development strategies based on affordability and risk. The system is considered a low development risk based on the use of terrestrial-derived reactor technology, high efficiency power conversion, and conventional materials. The low-risk approach was selected over other options that could offer higher performance and/or lower mass.

  13. EXPERIMENTAL LIQUID METAL FUEL REACTOR

    DOEpatents

    Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

    1962-01-23

    A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

  14. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    SciTech Connect

    Paul A. Demkowicz; David V. Laug; Dawn M. Scates; Edward L. Reber; Lyle G. Roybal; John B. Walter; Jason M. Harp; Robert N. Morris

    2012-10-01

    The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 degrees C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated fission gas monitoring system, as well as preliminary system calibration results.

  15. Uncertainties in nuclear fission data

    NASA Astrophysics Data System (ADS)

    Talou, Patrick; Kawano, Toshihiko; Chadwick, Mark B.; Neudecker, Denise; Rising, Michael E.

    2015-03-01

    We review the current status of our knowledge of nuclear fission data, and quantify uncertainties related to each fission observable whenever possible. We also discuss the roles that theory and experiment play in reducing those uncertainties, contributing to the improvement of our fundamental understanding of the nuclear fission process as well as of evaluated nuclear data libraries used in nuclear applications.

  16. Student Experiments in Spontaneous Fission.

    ERIC Educational Resources Information Center

    Becchetti, F. D.; Ying, J. S.

    1981-01-01

    Advanced undergraduate experiments utilizing a commercially available, thin spontaneous fission source are described, including studies of the energy and mass distribution of the fission fragments and their energy and angular correlation. The experiments provide a useful introduction to fission, nuclear mass equations, heavy-ion physics, and…

  17. Student Experiments in Spontaneous Fission.

    ERIC Educational Resources Information Center

    Becchetti, F. D.; Ying, J. S.

    1981-01-01

    Advanced undergraduate experiments utilizing a commercially available, thin spontaneous fission source are described, including studies of the energy and mass distribution of the fission fragments and their energy and angular correlation. The experiments provide a useful introduction to fission, nuclear mass equations, heavy-ion physics, and…

  18. REACTOR FUEL SCAVENGING MEANS

    DOEpatents

    Coffinberry, A.S.

    1962-04-10

    A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

  19. Causes, consequences, and kin bias of human group fissions.

    PubMed

    Walker, Robert S; Hill, Kim R

    2014-12-01

    Fissions of human communities are monumental occasions with consequences for cultural and genetic variation and divergence through time by means of serial founder effects. An ethnographic review shows that most human group fissions are fueled primarily by internal political conflict and secondarily by resource scarcity. As found for other social animals, human fissions lead to subgroups that have higher levels of relatedness as compared with the original community because of kin-biased assortment known as the lineal effect. Fission processes that increase the average relatedness of subgroups are important because relatedness governs how strongly kin/group selection favors social behaviors such as warfare, peacekeeping, and other forms of collection action. However, random individual assortment is not an appropriate null model for evaluating lineage assortment because nuclear families and extended households are expected to remain together, which in and of itself forces higher relatedness in smaller subgroups. We develop a lineage assortment index where low values represent subgroups with coefficients of relatedness near those expected if nuclear and extended households had chosen to associate into random groupings. Two fissions of Ache villages (Paraguay) are examples of this type of fission with a low lineage assortment index not significantly different from zero as evaluated with controlled simulations. On the other extreme, a lineage assortment index near unity represents a lineal fission that maximizes the relatedness of subgroups such as the perfect split of a lineage into sublineages. A fission of Piaroa (Venezuela) fits this scenario. While previous discussions of fission have emphasized similarities among human studies and even other social mammals, we highlight the full range of potential kin bias in the formation of new communities.

  20. Elastocapillary Instability in Mitochondrial Fission

    NASA Astrophysics Data System (ADS)

    Gonzalez-Rodriguez, David; Sart, Sébastien; Babataheri, Avin; Tareste, David; Barakat, Abdul I.; Clanet, Christophe; Husson, Julien

    2015-08-01

    Mitochondria are dynamic cell organelles that constantly undergo fission and fusion events. These dynamical processes, which tightly regulate mitochondrial morphology, are essential for cell physiology. Here we propose an elastocapillary mechanical instability as a mechanism for mitochondrial fission. We experimentally induce mitochondrial fission by rupturing the cell's plasma membrane. We present a stability analysis that successfully explains the observed fission wavelength and the role of mitochondrial morphology in the occurrence of fission events. Our results show that the laws of fluid mechanics can describe mitochondrial morphology and dynamics.

  1. 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.

  2. Membrane fission by protein crowding.

    PubMed

    Snead, Wilton T; Hayden, Carl C; Gadok, Avinash K; Zhao, Chi; Lafer, Eileen M; Rangamani, Padmini; Stachowiak, Jeanne C

    2017-04-18

    Membrane fission, which facilitates compartmentalization of biological processes into discrete, membrane-bound volumes, is essential for cellular life. Proteins with specific structural features including constricting rings, helical scaffolds, and hydrophobic membrane insertions are thought to be the primary drivers of fission. In contrast, here we report a mechanism of fission that is independent of protein structure-steric pressure among membrane-bound proteins. In particular, random collisions among crowded proteins generate substantial pressure, which if unbalanced on the opposite membrane surface can dramatically increase membrane curvature, leading to fission. Using the endocytic protein epsin1 N-terminal homology domain (ENTH), previously thought to drive fission by hydrophobic insertion, our results show that membrane coverage correlates equally with fission regardless of the hydrophobicity of insertions. Specifically, combining FRET-based measurements of membrane coverage with multiple, independent measurements of membrane vesiculation revealed that fission became spontaneous as steric pressure increased. Further, fission efficiency remained equally potent when helices were replaced by synthetic membrane-binding motifs. These data challenge the view that hydrophobic insertions drive membrane fission, suggesting instead that the role of insertions is to anchor proteins strongly to membrane surfaces, amplifying steric pressure. In line with these conclusions, even green fluorescent protein (GFP) was able to drive fission efficiently when bound to the membrane at high coverage. Our conclusions are further strengthened by the finding that intrinsically disordered proteins, which have large hydrodynamic radii yet lack a defined structure, drove fission with substantially greater potency than smaller, structured proteins.

  3. Spent fuel data for waste storage programs

    SciTech Connect

    Greene, E M

    1980-09-01

    Data on LWR spent fuel were compiled for dissemination to participants in DOE-sponsored waste storage programs. Included are mechanical descriptions of the existing major types of LWR fuel assemblies, spent LWR fuel fission product inventories and decay heat data, and inventories of LWR spent fuel currently in storage, with projections of future quantities.

  4. JACKETED FISSIONABLE MEMBER

    DOEpatents

    Boller, E.R.; Robinson, J.W.

    1960-09-13

    A fuel element design for a nuclear reactor is presented. The fuel element comprises a cylindrical fuel body having a portion of smaller diameter at each end thereof with an annular flange at the extreme ends of these portions of smaller diameter. An end cap fits over the ends of the fuel body and has an internal annular groove adapted to receive the flange. The fuel body and end caps are disposed in a cup-shaped jacket, a closure disc completing the enclosure of the fuel body, and tht caps are bonded over their entire periphery to the jacket.

  5. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-01

    A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

  6. Modeling of Fission Gas Release in UO2

    SciTech Connect

    MH Krohn

    2006-01-23

    A two-stage gas release model was examined to determine if it could provide a physically realistic and accurate model for fission gas release under Prometheus conditions. The single-stage Booth model [1], which is often used to calculate fission gas release, is considered to be oversimplified and not representative of the mechanisms that occur during fission gas release. Two-stage gas release models require saturation at the grain boundaries before gas is release, leading to a time delay in release of gases generated in the fuel. Two versions of a two-stage model developed by Forsberg and Massih [2] were implemented using Mathcad [3]. The original Forsbers and Massih model [2] and a modified version of the Forsberg and Massih model that is used in a commercially available fuel performance code (FRAPCON-3) [4] were examined. After an examination of these models, it is apparent that without further development and validation neither of these models should be used to calculate fission gas release under Prometheus-type conditions. There is too much uncertainty in the input parameters used in the models. In addition. the data used to tune the modified Forsberg and Massih model (FRAPCON-3) was collected under commercial reactor conditions, which will have higher fission rates relative to Prometheus conditions [4].

  7. Space Fission Propulsion Testing and Development Progress. Phase 1

    NASA Technical Reports Server (NTRS)

    VanDyke, Melissa; Houts, Mike; Pedersen, Kevin; Godfroy, Tom; Dickens, Ricky; Poston, David; Reid, Bob; Salvail, Pat; Ring, Peter; Rodgers, Stephen L. (Technical Monitor)

    2001-01-01

    Successful development of space fission systems will require an extensive program of affordable and realistic testing. In addition to tests related to design/development of the fission system, realistic testing of the actual flight unit must also be performed. Testing can be divided into two categories, non-nuclear tests and nuclear tests. Full power nuclear tests of space fission systems we expensive, time consuming, and of limited use, even in the best of programmatic environments. If the system is designed to operate within established radiation damage and fuel burn up limits while simultaneously being designed to allow close simulation of heat from fission using resistance heaters, high confidence in fission system performance and lifetime can be attained through a series of non-nuclear tests. Non-nuclear tests are affordable and timely, and the cause of component and system failures can be quickly and accurately identified. MSFC is leading a Safe Affordable Fission Engine (SAFE) test series whose ultimate goal is the demonstration of a 300 kW flight configuration system using non-nuclear testing. This test series is carried out in collaboration with other NASA centers, other government agencies, industry, and universities. If SAFE-related nuclear tests are desired they will have a high probability of success and can be performed at existing nuclear facilities. The paper describes the SAFE non-nuclear test series, which includes test article descriptions, test results and conclusions, and future test plans.

  8. Phase 1 space fission propulsion system testing and development progress

    NASA Astrophysics Data System (ADS)

    van Dyke, Melissa; Houts, Mike; Pedersen, Kevin; Godfroy, Tom; Dickens, Ricky; Poston, David; Reid, Bob; Salvail, Pat; Ring, Peter

    2001-02-01

    Successful development of space fission systems will require an extensive program of affordable and realistic testing. In addition to tests related to design/development of the fission system, realistic testing of the actual flight unit must also be performed. Testing can be divided into two categories, non-nuclear tests and nuclear tests. Full power nuclear tests of space fission systems are expensive, time consuming, and of limited use, even in the best of programmatic environments. If the system is designed to operate within established radiation damage and fuel burn up limits while simultaneously being designed to allow close simulation of heat from fission using resistance heaters, high confidence in fission system performance and lifetime can be attained through a series of non-nuclear tests. Non-nuclear tests are affordable and timely, and the cause of component and system failures can be quickly and accurately identified, MSFC is leading a Safe Affordable Fission Engine (SAFE) test series whose ultimate goal is the demonstration of a 300 kW flight configuration system using non-nuclear testing. This test series is carried out in collaboration with other NASA centers, other government agencies, industry, and universities. If SAFE-related nuclear tests are desired, they will have a high probability of success and can be performed at existing nuclear facilities. The paper describes the SAFE non-nuclear test series, which includes test article descriptions, test results and conclusions, and future test plans. .

  9. Fission Yields and Other Diagnostics for Nuclear Performance

    SciTech Connect

    Chadwick, M.B.

    2014-06-15

    I summarize advances in our understanding of basic nuclear physics cross sections and decay properties that are needed to characterize the magnitude and energy-dependence of a neutron flux, and to determine the amount of fission burnup in plutonium fuel. The number of fissions that have occurred in a neutron environment can be deduced from measurements of the fission products created, providing that the fission product yields are known accurately. I describe how our understanding of plutonium fission product yields has improved in recent years through a meta-analysis of various measured data, and through identification of fission product yield incident-energy dependencies over the 0.2-2 MeV fast energy region. This led to the resolution of a previous discrepancy between the Los Alamos and Lawrence Livermore National Laboratories in their plutonium yield assessments in the fast energy region, although more experimental work is still needed to resolve discrepancies at 14 MeV. Work is also described that has improved our understanding of (n,2n) cross sections that are used as diagnostics of the high-energy neutron spectrum – both on plutonium and americium, and on the radiochemical detectors yttrium, iridium, and thulium. Finally, some observations are made on the importance of continuing to develop our Evaluated Nuclear Data Files (ENDF) database using physics insights from differential cross section and integral laboratory experiments and from nuclear theory advances.

  10. Fission Yields and Other Diagnostics for Nuclear Performance

    NASA Astrophysics Data System (ADS)

    Chadwick, M. B.

    2014-06-01

    I summarize advances in our understanding of basic nuclear physics cross sections and decay properties that are needed to characterize the magnitude and energy-dependence of a neutron flux, and to determine the amount of fission burnup in plutonium fuel. The number of fissions that have occurred in a neutron environment can be deduced from measurements of the fission products created, providing that the fission product yields are known accurately. I describe how our understanding of plutonium fission product yields has improved in recent years through a meta-analysis of various measured data, and through identification of fission product yield incident-energy dependencies over the 0.2-2 MeV fast energy region. This led to the resolution of a previous discrepancy between the Los Alamos and Lawrence Livermore National Laboratories in their plutonium yield assessments in the fast energy region, although more experimental work is still needed to resolve discrepancies at 14 MeV. Work is also described that has improved our understanding of (n,2n) cross sections that are used as diagnostics of the high-energy neutron spectrum - both on plutonium and americium, and on the radiochemical detectors yttrium, iridium, and thulium. Finally, some observations are made on the importance of continuing to develop our Evaluated Nuclear Data Files (ENDF) database using physics insights from differential cross section and integral laboratory experiments and from nuclear theory advances.

  11. Data summary report for fission product release test VI-6

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Travis, J.R.; Webster, C.S.; Collins, J.L.

    1994-03-01

    Test VI-6 was the sixth test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium. The fuel had experienced a burnup of {approximately}42 MWd/kg, with inert gas release during irradiation of {approximately}2%. The fuel specimen was heated in an induction furnace at 2300 K for 60 min, initially in hydrogen, then in a steam atmosphere. The released fission products were collected in three sequentially operated collection trains designed to facilitate sampling and analysis. The fission product inventories in the fuel were measured directly by gamma-ray spectrometry, where possible, and were calculated by ORIGEN2. Integral releases were 75% for {sup 85}Kr, 67% for {sup 129}I, 64% for {sup 125}Sb, 80% for both {sup 134}Cs and {sup 137}Cs, 14% for {sup 154}Eu, 63% for Te, 32% for Ba, 13% for Mo, and 5.8% for Sr. Of the totals released from the fuel, 43% of the Cs, 32% of the Sb, and 98% of the Eu were deposited in the outlet end of the furnace. During the heatup in hydrogen, the Zircaloy cladding melted, ran down, and reacted with some of the UO{sub 2} and fission products, especially Te and Sb. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.57 g, almost equally divided between thermal gradient tubes and filters. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL Diffusion Model.

  12. Fission-Fusion Neutron Source Progress Report Sept 30, 2009

    SciTech Connect

    Chapline, G F; Daffin, F; Clark, R

    2010-02-19

    In this report the authors describe the progress made in FY09 in evaluating the feasibility of a new concept for using the DT fusion reaction to produce intense pulses of 14 MeV neutrons. In this new scheme the heating of the DT is accomplished using fission fragments rather than ion beams as in conventional magnet confinement fusion schemes or lasers in inertial confinement schemes. As a source of fission fragments they propose using a dust reactor concept introduced some time ago by one of us (RC). An attractive feature of this approach is that there is no need for a large auxiliary power source to heat the DT plasma to the point where self-sustaining fusion become possible. Their scheme does require pulsed magnetic fields, but generating these fields requires only a modest power source. The dust reactor that they propose using for their neutron source would use micron-sized UC pellets suspended in a vacuum as the reactor fuel. Surrounding the fuel with a moderator such as heavy water (D{sub 2}O) would allow the reactor to operate as a thermal reactor and require only modest amounts of HEU. The scheme for using fission fragments to generate intense pulses of 14 MeV neutrons is based on the fission fragment rocket idea. In the fission fragment rocket scheme it was contemplated that the fission fragments produced in a low density reactor core could be guided out of the reactor by large magnetic fields used to form a 'rocket exhaust'. Their adaptation of this idea for the purposes of making a neutron source involves using the fission fragments escaping from one side of a tandem magnet mirror to heat DT gas confined in the adjacent magnetic trap.

  13. Safety characteristics of a suspended-pellet fission reactor system

    NASA Astrophysics Data System (ADS)

    Kingdon, David Ross

    A new fission reactor system with passive safety characteristics to eliminate the occurrence of loss-of-coolant accidents, reduce reactivity excursion effects, and which also provides for closure of the nuclear fuel cycle through on-site spent fuel management is examined. The concept uses multi-coated fuel pellets which are suspended by an upward moving coolant in vertical columns of the reactor core and electro-refining elemental separation to remove selected fission products prior to actinide recycling. The possibility of fuel melt following a loss-of-coolant is avoided as a decrease in coolant flow results in the removal of fuel from the core through the action of gravity alone. Average fluid velocities in the columns which are necessary to suspend the pellets are calculated and found to be consistent with the necessary heat extraction to yield ˜1--10 Wth per column. The total output power of such suspended pellet-type reactors is compared to the power necessary to provide the suspending fluid flow, yielding favourable ratios of ˜102--103. The reduction of reactivity excursion tendencies is envisaged through an ablative layer of material in the pellets which sublimates at temperatures above normal operating conditions. In the event of a power or temperature increase the particles fragment and thereby change their hydrodynamic drag characteristics, thus leading to fuel removal from the core by elutriation. Comparison of nuclear-to-thermal response times and elutriation rates for limiting power transients indicate that the present design assists in reactivity excursion mitigation. Closure of the nuclear fuel cycle is attained through a spent fuel management strategy which requires only on-site storage of a fraction of the fission products produced during reactor operation. Electro-refining separation of selected fission products combined with complete actinide recycling yields no isolation of plutonium or highly enriched uranium during the procedure. The out

  14. Extended optical model for fission

    DOE PAGES

    Sin, M.; Capote, R.; Herman, M. W.; ...

    2016-03-07

    A comprehensive formalism to calculate fission cross sections based on the extension of the optical model for fission is presented. It can be used for description of nuclear reactions on actinides featuring multi-humped fission barriers with partial absorption in the wells and direct transmission through discrete and continuum fission channels. The formalism describes the gross fluctuations observed in the fission probability due to vibrational resonances, and can be easily implemented in existing statistical reaction model codes. The extended optical model for fission is applied for neutron induced fission cross-section calculations on 234,235,238U and 239Pu targets. A triple-humped fission barrier ismore » used for 234,235U(n,f), while a double-humped fission barrier is used for 238U(n,f) and 239Pu(n,f) reactions as predicted by theoretical barrier calculations. The impact of partial damping of class-II/III states, and of direct transmission through discrete and continuum fission channels, is shown to be critical for a proper description of the measured fission cross sections for 234,235,238U(n,f) reactions. The 239Pu(n,f) reaction can be calculated in the complete damping approximation. Calculated cross sections for 235,238U(n,f) and 239Pu(n,f) reactions agree within 3% with the corresponding cross sections derived within the Neutron Standards least-squares fit of available experimental data. Lastly, the extended optical model for fission can be used for both theoretical fission studies and nuclear data evaluation.« less

  15. Extended optical model for fission

    SciTech Connect

    Sin, M.; Capote, R.; Herman, M. W.; Trkov, A.

    2016-03-07

    A comprehensive formalism to calculate fission cross sections based on the extension of the optical model for fission is presented. It can be used for description of nuclear reactions on actinides featuring multi-humped fission barriers with partial absorption in the wells and direct transmission through discrete and continuum fission channels. The formalism describes the gross fluctuations observed in the fission probability due to vibrational resonances, and can be easily implemented in existing statistical reaction model codes. The extended optical model for fission is applied for neutron induced fission cross-section calculations on 234,235,238U and 239Pu targets. A triple-humped fission barrier is used for 234,235U(n,f), while a double-humped fission barrier is used for 238U(n,f) and 239Pu(n,f) reactions as predicted by theoretical barrier calculations. The impact of partial damping of class-II/III states, and of direct transmission through discrete and continuum fission channels, is shown to be critical for a proper description of the measured fission cross sections for 234,235,238U(n,f) reactions. The 239Pu(n,f) reaction can be calculated in the complete damping approximation. Calculated cross sections for 235,238U(n,f) and 239Pu(n,f) reactions agree within 3% with the corresponding cross sections derived within the Neutron Standards least-squares fit of available experimental data. Lastly, the extended optical model for fission can be used for both theoretical fission studies and nuclear data evaluation.

  16. Extended optical model for fission

    NASA Astrophysics Data System (ADS)

    Sin, M.; Capote, R.; Herman, M. W.; Trkov, A.

    2016-03-01

    A comprehensive formalism to calculate fission cross sections based on the extension of the optical model for fission is presented. It can be used for description of nuclear reactions on actinides featuring multi-humped fission barriers with partial absorption in the wells and direct transmission through discrete and continuum fission channels. The formalism describes the gross fluctuations observed in the fission probability due to vibrational resonances, and can be easily implemented in existing statistical reaction model codes. The extended optical model for fission is applied for neutron induced fission cross-section calculations on 234,235,238U and 239Pu targets. A triple-humped fission barrier is used for U,235234(n ,f ) , while a double-humped fission barrier is used for 238U(n ,f ) and 239Pu(n ,f ) reactions as predicted by theoretical barrier calculations. The impact of partial damping of class-II/III states, and of direct transmission through discrete and continuum fission channels, is shown to be critical for a proper description of the measured fission cross sections for 234,235,238U(n ,f ) reactions. The 239Pu(n ,f ) reaction can be calculated in the complete damping approximation. Calculated cross sections for U,238235(n ,f ) and 239Pu(n ,f ) reactions agree within 3% with the corresponding cross sections derived within the Neutron Standards least-squares fit of available experimental data. The extended optical model for fission can be used for both theoretical fission studies and nuclear data evaluation.

  17. A model for nonvolatile fission product release during reactor accident conditions

    SciTech Connect

    Lewis, B.J.; Andre, B.; Ducros, G.; Maro, D.

    1996-10-01

    An analytical model has been developed to describe the release kinetics of nonvolatile fission products (e.g., molybdenum, cerium, ruthenium, and barium) from uranium dioxide fuel under severe reactor accident conditions. This treatment considers the rate-controlling process of release in accordance with diffusional transport in the fuel matrix and fission product vaporization from the fuel surface into the surrounding gas atmosphere. The effect of the oxygen potential in the gas atmosphere on the chemical form and volatility of the fission product is considered. A correlation is also developed to account for the trapping effects of antimony and tellurium in the Zircaloy cladding. This model interprets the release behavior of fission products observed in Commissariat a l`Energie Atomique experiments conducted in the HEVA/VERCORS facility at high temperature in a hydrogen and steam atmosphere.

  18. Metallic fuel development

    SciTech Connect

    Walters, L.C.

    1987-01-01

    Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising.

  19. Early Flight Fission Test Facilities (EFF-TF) and Concepts That Support Near-Term Space Fission Missions

    NASA Technical Reports Server (NTRS)

    VanDyke, Melissa; Houts, Mike; Godfroy, Thomas; Martin, James

    2003-01-01

    Fission technology can enable rapid, affordable access to any point in the solar system. If fusion propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and utilized. Successful utilization will most likely occur if frequent, significant hardware-based milestones can be achieved throughout the program. If the system is designed to operate within established radiation damage and fuel burn up limits while simultaneously being designed to allow close simulation of heat from fission using resistance heaters, high confidence in fission system pe$ormance and lifetime can be attained through non-nuclear testing. Through demonstration of systems concepts (designed by DOE National Laboratories) in relevant environments, this philosophy has been demonstrated through hardware testing in the Early Flight Fission Test Facilities (EFF-TF) at the Marshall Space Flight Center. The EFF-TF is designed to enable very realistic non-nuclear testing of space fission systems. Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE labs, industry, universities, and other NASA centers.

  20. Fission-induced plasmas

    NASA Technical Reports Server (NTRS)

    Harries, W. L.; Shiu, Y. J.

    1979-01-01

    The possibility of creating a plasma from fission fragments, and to utilize the energy of the particles to create population inversion that would lead to laser action is investigated. An investigation was made of various laser materials which could be used for nuclear-pumped lasing. The most likely candidate for a fissioning material in the gaseous form is uranium hexafluoride - UF6, and experiments were performed to investigate materials that would be compatible with it. One of the central problems in understanding a fission-induced plasma is to obtain a model of the electron behavior, and some preliminary calculations are presented. In particular, the rates of various processes are discussed. A simple intuitive model of the electron energy distribution function is also shown. The results were useful for considering a mathematical model of a nuclear-pumped laser. Next a theoretical model of a (3)He-Ar nuclear-pumped laser is presented. The theory showed good qualitative agreement with the experimental results.

  1. Axisymmetric Magnetic Mirror Fusion-Fission Hybrid

    SciTech Connect

    Moir, R. W.; Martovetsky, N. N.; Molvik, A. W.; Ryutov, D. D.; Simonen, T. C.

    2011-05-13

    sale of both electricity and production of fissile fuel with fissioning blankets might need Q~1 while suppressing fissioning might be the most economical application of fusion but will require Q≥4.

  2. PolyPole-1: An accurate numerical algorithm for intra-granular fission gas release

    SciTech Connect

    Pizzocri, D.; Rabiti, C.; Luzzi, L.; Barani, T.; Van Uffelen, P.; Pastore, G.

    2016-09-01

    This paper describes the development of a new numerical algorithm (called PolyPole-1) to efficiently solve the equation for intra-granular fission gas release in nuclear fuel. The work was carried out in collaboration with Politecnico di Milano and Institute for Transuranium Elements. The PolyPole-1 algorithms is being implemented in INL's fuels code BISON code as part of BISON's fission gas release model. The transport of fission gas from within the fuel grains to the grain boundaries (intra-granular fission gas release) is a fundamental controlling mechanism of fission gas release and gaseous swelling in nuclear fuel. Hence, accurate numerical solution of the corresponding mathematical problem needs to be included in fission gas behaviour models used in fuel performance codes. Under the assumption of equilibrium between trapping and resolution, the process can be described mathematically by a single diffusion equation for the gas atom concentration in a grain. In this work, we propose a new numerical algorithm (PolyPole-1) to efficiently solve the fission gas diffusion equation in time-varying conditions. The PolyPole-1 algorithm is based on the analytic modal solution of the diffusion equation for constant conditions, with the addition of polynomial corrective terms that embody the information on the deviation from constant conditions. The new algorithm is verified by comparing the results to a finite difference solution over a large number of randomly generated operation histories. Furthermore, comparison to state-of-the-art algorithms used in fuel performance codes demonstrates that the accuracy of the PolyPole-1 solution is superior to other algorithms, with similar computational effort. Finally, the concept of PolyPole-1 may be extended to the solution of the general problem of intra-granular fission gas diffusion during non-equilibrium trapping and resolution, which will be the subject of future work.

  3. Fission-Based Electric Propulsion for Interstellar Precursor Missions

    SciTech Connect

    HOUTS,MICHAEL G.; LENARD,ROGER X.; LIPINSKI,RONALD J.; PATTON,BRUCE; POSTON,DAVID; WRIGHT,STEVEN A.

    1999-11-03

    This paper reviews the technology options for a fission-based electric propulsion system for interstellar precursor missions. To achieve a total {Delta}V of more than 100 km/s in less than a decade of thrusting with an electric propulsion system of 10,000s Isp requires a specific mass for the power system of less than 35 kg/kWe. Three possible configurations are described: (1) a UZrH-fueled,NaK-cooled reactor with a steam Rankine conversion system,(2) a UN-fueled gas-cooled reactor with a recuperated Brayton conversion system, and (3) a UN-fueled heat pipe-cooled reactor with a recuperated Brayton conversion system. All three of these systems have the potential to meet the specific mass requirements for interstellar precursor missions in the near term. Advanced versions of a fission-based electric propulsion system might travel as much as several light years in 200 years.

  4. Chemical state of fission products in irradiated UO 2

    NASA Astrophysics Data System (ADS)

    Imoto, S.

    1986-08-01

    The chemical state of fission products in irradiated UO 2 fuel has been estimated for FBR as well as LWR on the basis of equilibrium calculation with the SOLGASMIX-PV code. The system considered for the calculation is composed of a gas phase, a CaF 2 type oxide phase, three grey phases, a noble metal alloy, a mixed telluride phase and several other phases each consisting of single compound. The distribution of elements into these phases and the amount of chemical species in each phase at different temperatures are obtained as a function of oxygen potential for LWR and FBR. Changes of the chemical potential of the fuel-fission products system during burnup are also evaluated with particular attention to the difference between LWR and FBR. Some informations obtained by the calculation are compared with the results of post irradiation examination of UO 2 fuels.

  5. Report on simulation of fission gas and fission product diffusion in UO2

    SciTech Connect

    Andersson, Anders David; Perriot, Romain Thibault; Pastore, Giovanni; Tonks, Michael R.; Cooper, Michael William; Liu, Xiang-Yang; Goyal, Anuj; Uberuaga, Blas P.; Stanek, Christopher Richard

    2016-07-22

    In UO2 nuclear fuel, the retention and release of fission gas atoms such as xenon (Xe) are important for nuclear fuel performance by, for example, reducing the fuel thermal conductivity, causing fuel swelling that leads to mechanical interaction with the clad, increasing the plenum pressure and reducing the fuel–clad gap thermal conductivity. We use multi-­scale simulations to determine fission gas diffusion mechanisms as well as the corresponding rates in UO2 under both intrinsic and irradiation conditions. In addition to Xe and Kr, the fission products Zr, Ru, Ce, Y, La, Sr and Ba have been investigated. Density functional theory (DFT) calculations are used to study formation, binding and migration energies of small clusters of Xe atoms and vacancies. Empirical potential calculations enable us to determine the corresponding entropies and attempt frequencies for migration as well as investigate the properties of large clusters or small fission gas bubbles. A continuum reaction-­diffusion model is developed for Xe and point defects based on the mechanisms and rates obtained from atomistic simulations. Effective fission gas diffusivities are then obtained by solving this set of equations for different chemical and irradiation conditions using the MARMOT phase field code. The predictions are compared to available experimental data. The importance of the large XeU3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence of its high mobility and high binding energy. We find that the XeU3O cluster gives Xe diffusion coefficients that are higher for intrinsic conditions than under irradiation over a wide range of temperatures. Under irradiation the fast-­moving XeU3O cluster recombines quickly with irradiation-induced interstitial U ions, while this mechanism is less important for intrinsic conditions. The net result is higher

  6. FUEL ELEMENT FOR A NEUTRONIC REACTOR

    DOEpatents

    Tonks, L.

    1959-09-22

    A fuel element is presented for a reactor comprising a stack of conical bodies of nonfissionable material disposed with the vertex up carrying wafers of fissionable material in grooves near the periphery. These bodies are in a jacket which contains a thermally conducting liquid immersing the bodies. Gaseous fission preducts pass upwardly through central apertures in the bodies while fragments of fissionable material are trapped by vertical projections or walls on the upper surface of the bodies.

  7. NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

    DOEpatents

    Rasor, N.S.; Hirsch, R.L.

    1963-12-01

    The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

  8. The Radiological and Thermal Characteristics of Fission Waste from a Deep-Burn Fusion-Fission Hybrid (LIFE) and Implications for Repository Performance

    SciTech Connect

    Shaw, H F; Blink, J; Farmer, J; Latkowski, J; Kramer, K

    2009-09-08

    We are studying the use of a Laser Inertial-confinement Fusion Engine (LIFE) to drive a hybrid fusion-fission system that can generate electrical power and/or burn nuclear waste. The system uses the neutrons from laser driven ICF to produce tritium and to drive nuclear reactions in a subcritical fission blanket. The fusion neutron source obviates the need for a self-sustaining chain reaction in the fission blanket. Either fissile or fertile could be used as fission fuel, thus eliminating the need for isotopic enrichment. The 'driven' system potentially allows very high levels of burnup to be reached, extracting a large fraction of the available energy in the fission fuel without the need for reprocessing. In this note, we discuss the radionuclide inventory of a depleted uranium (DU) fuel burned to greater than 95% FIMA (Fissions per Initial heavy Metal Atom), the implications for thermal management of the resulting waste, and the implications of this waste for meeting the dose standards for releases from a geological repository for high-level waste. The fission waste discussed here would be that produced by a LIFE hybrid with a 500-MW fusion source. The fusion neutrons are multiplied and moderated by a sequence of concentric shells of materials before encountering the fission fuel, and fission in this region is largely due to thermal neutrons. The fission blanket consists of 40 metric tons (MT) of DU, assumed to be in the form of TRISO-like UOC fuel particles embedded in 2-cm-diameter graphite pebbles. (It is recognized that TRISO-based fuel may not reach the high burnup of the fertile fuel considered here, and other fuel options are being investigated. We postulate the existence of a fuel that can reach >95% FIMA so that the waste disposal implications of high burnup can be assessed.) The engine and plant design considered here would receive one load of fission fuel and produce {approx}2 GWt of power (fusion + fission) over its 50- to 70-year lifetime. Neutron and

  9. Fission yield measurements at IGISOL

    NASA Astrophysics Data System (ADS)

    Lantz, M.; Al-Adili, A.; Gorelov, D.; Jokinen, A.; Kolhinen, V. S.; Mattera, A.; Moore, I.; Penttilä, H.; Pomp, S.; Prokofiev, A. V.; Rakopoulos, V.; Rinta-Antila, S.; Simutkin, V.; Solders, A.

    2016-06-01

    The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL) technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f) and Th(p,f) have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn) reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

  10. Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on {sup 239}Pu, {sup 235}U, {sup 238}U

    SciTech Connect

    Selby, H.D.; Mac Innes, M.R.; Barr, D.W.; Keksis, A.L.; Meade, R.A.; Burns, C.J.; Chadwick, M.B.; Wallstrom, T.C.

    2010-12-15

    We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for {sup 99}Mo, {sup 95}Zr, {sup 137}Cs, {sup 140}Ba, {sup 141,143}Ce, and {sup 147}Nd. Modest incident-energy dependence exists for the {sup 147}Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by {approx}5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried

  11. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  12. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and

  13. Fission fragment charge and mass distributions in 239Pu(n, f ) in the adiabatic nuclear energy density functional theory

    DOE PAGES

    Regnier, D.; Dubray, N.; Schunck, N.; ...

    2016-05-13

    Here, accurate knowledge of fission fragment yields is an essential ingredient of numerous applications ranging from the formation of elements in the r process to fuel cycle optimization for nuclear energy. The need for a predictive theory applicable where no data are available, together with the variety of potential applications, is an incentive to develop a fully microscopic approach to fission dynamics.

  14. The SPIDER fission fragment spectrometer for fission product yield measurements

    SciTech Connect

    Meierbachtol, K.; Tovesson, F.; Shields, D.; Arnold, C.; Blakeley, R.; Bredeweg, T.; Devlin, M.; Hecht, A. A.; Heffern, L. E.; Jorgenson, J.; Laptev, A.; Mader, D.; O׳Donnell, J. M.; Sierk, A.; White, M.

    2015-04-01

    We developed the SPectrometer for Ion DEtermination in fission Research (SPIDER) for measuring mass yield distributions of fission products from spontaneous and neutron-induced fission. The 2E–2v method of measuring the kinetic energy (E) and velocity (v) of both outgoing fission products has been utilized, with the goal of measuring the mass of the fission products with an average resolution of 1 atomic mass unit (amu). Moreover, the SPIDER instrument, consisting of detector components for time-of-flight, trajectory, and energy measurements, has been assembled and tested using 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission products from spontaneous fission of 252Cf. Individual measurement resolutions were met for time-of-flight (250 ps FWHM), spacial resolution (2 mm FHWM), and energy (92 keV FWHM for 8.376 MeV). Finally, these mass yield results measured from 252Cf spontaneous fission products are reported from an E–v measurement.

  15. The SPIDER fission fragment spectrometer for fission product yield measurements

    DOE PAGES

    Meierbachtol, K.; Tovesson, F.; Shields, D.; ...

    2015-04-01

    We developed the SPectrometer for Ion DEtermination in fission Research (SPIDER) for measuring mass yield distributions of fission products from spontaneous and neutron-induced fission. The 2E–2v method of measuring the kinetic energy (E) and velocity (v) of both outgoing fission products has been utilized, with the goal of measuring the mass of the fission products with an average resolution of 1 atomic mass unit (amu). Moreover, the SPIDER instrument, consisting of detector components for time-of-flight, trajectory, and energy measurements, has been assembled and tested using 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission productsmore » from spontaneous fission of 252Cf. Individual measurement resolutions were met for time-of-flight (250 ps FWHM), spacial resolution (2 mm FHWM), and energy (92 keV FWHM for 8.376 MeV). Finally, these mass yield results measured from 252Cf spontaneous fission products are reported from an E–v measurement.« less

  16. Measurement of Fission Product Yields from Fast-Neutron Fission

    NASA Astrophysics Data System (ADS)

    Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

    2014-09-01

    One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

  17. Post-irradiation Examination and Fission Product Inventory Analysis of AGR-1 Irradiation Capsules

    SciTech Connect

    J M Harp; P D Demkowicz; S A Ploger

    2012-10-01

    The AGR-1 experiment was the first in a series of Advanced Gas Reactor (AGR) experiments designed to test TRISO fuel under High Temperature Gas Reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL’s Materials and Fuels Complex (MFC). The inventory and distribution of fission products, especially Ag-110m, was assessed and analyzed for all the components of the AGR-1 capsules. This data should help inform the study of fission product migration in coated particle fuel. Gamma spectrometry was used to measure the activity of various different fission products in the different components of the AGR-1 test train. Each capsule contained: 12 fuel compacts, a graphite holder that kept the fuel compacts in place, graphite spacers that were above and below the graphite holders and fuel compacts, gas lines through which a helium neon gas mixture flowed in and out of each capsule, and the stainless steel shell that contained the experiment. Gamma spectrometry results and the experimental techniques used to capture these results will be presented for all the capsule components. The components were assayed to determine the total activity of different fission products present in or on them. These totals are compared to the total expected activity of a particular fission product in the capsule based on predictions from physics simulation. Based on this metric, a significant fraction of the Ag-110m was detected outside the fuel compacts, but the amount varied highly between the 6 capsules. Very small fractions of Cs-137 (<2E-5), Cs-134 (<1e-5), and Eu-154 (<4e-4) were detected outside of the fuel compacts. Additionally, the distribution of select fission products in some of the components including the fuel compacts and the graphite holders were measured and will be discussed.

  18. Fission: The first 50 years

    SciTech Connect

    Vandenbosch, R.

    1989-01-01

    The possibility of fission had been largely unanticipated prior to its discovery in 1938. This process, with its dramatically large energy release and its formation of previously unknown nuclides, immediately captured the imagination of the scientific community. Both theoretical and experimental developments occurred at a rapid pace. I will begin my discussion of fission with the far-reaching paper of Bohr and Wheeler, who in little more than half a year laid out a framework for understanding many features of the fission process. I will then turn to our current understanding of a number of aspects of fission. One of these is the pronounced tendency of many nuclear species to fission asymmetrically. In fact, the discovery of fission was based on the identification of barium isotopes produced in asymmetric fission. The dramatic changes in the preferred mass division and kinetic energy release with the addition of only a few neutrons to the spontaneously fissioning Fermium isotopes will be emphasized. The problem of the dynamics of saddle to scission will be discussed---this is one aspect of fission for which we do not have all the answers. Another dynamical effect to be discussed is the apparent failure of transition state theory at high excitation energies. The role of single particle (shell) effects in enriching the structure if the potential energy surface will be explored. Spontaneously fissioning isomers and intermediate structure resonances will be discussed. The recognition that short-lived fission isomers are superdeformed shape isomers has been followed by the recent observation of superdeformed shape isomers in the rare earth region. 18 refs., 3 figs.

  19. Comparison of Fission Product Yields and Their Impact

    SciTech Connect

    S. Harrison

    2006-02-01

    This memorandum describes the Naval Reactors Prime Contractor Team (NRPCT) Space Nuclear Power Program (SNPP) interest in determining the expected fission product yields from a Prometheus-type reactor and assessing the impact of these species on materials found in the fuel element and balance of plant. Theoretical yield calculations using ORIGEN-S and RACER computer models are included in graphical and tabular form in Attachment, with focus on the desired fast neutron spectrum data. The known fission product interaction concerns are the corrosive attack of iron- and nickel-based alloys by volatile fission products, such as cesium, tellurium, and iodine, and the radiological transmutation of krypton-85 in the coolant to rubidium-85, a potentially corrosive agent to the coolant system metal piping.

  20. Precise Nuclear Data Measurements Possible with the NIFFTE fissionTPC for Advanced Reactor Designs

    NASA Astrophysics Data System (ADS)

    Towell, Rusty; Niffte Collaboration

    2015-10-01

    The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) Collaboration has applied the proven technology of Time Projection Chambers (TPC) to the task of precisely measuring fission cross sections. With the NIFFTE fission TPC, precise measurements have been made during the last year at the Los Alamos Neutron Science Center from both U-235 and Pu-239 targets. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow systematics to be controlled at the level of 1%. The fissionTPC performance will be presented. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors such as Liquid Fluoride Thorium Reactors over uranium-fueled reactors. These advantages include improved reactor safety, minimizing radioactive waste, improved reactor efficiency, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium-fueled reactors will also be discussed.

  1. PolyPole-1: An accurate numerical algorithm for intra-granular fission gas release

    NASA Astrophysics Data System (ADS)

    Pizzocri, D.; Rabiti, C.; Luzzi, L.; Barani, T.; Van Uffelen, P.; Pastore, G.

    2016-09-01

    The transport of fission gas from within the fuel grains to the grain boundaries (intra-granular fission gas release) is a fundamental controlling mechanism of fission gas release and gaseous swelling in nuclear fuel. Hence, accurate numerical solution of the corresponding mathematical problem needs to be included in fission gas behaviour models used in fuel performance codes. Under the assumption of equilibrium between trapping and resolution, the process can be described mathematically by a single diffusion equation for the gas atom concentration in a grain. In this paper, we propose a new numerical algorithm (PolyPole-1) to efficiently solve the fission gas diffusion equation in time-varying conditions. The PolyPole-1 algorithm is based on the analytic modal solution of the diffusion equation for constant conditions, combined with polynomial corrective terms that embody the information on the deviation from constant conditions. The new algorithm is verified by comparing the results to a finite difference solution over a large number of randomly generated operation histories. Furthermore, comparison to state-of-the-art algorithms used in fuel performance codes demonstrates that the accuracy of PolyPole-1 is superior to other algorithms, with similar computational effort. Finally, the concept of PolyPole-1 may be extended to the solution of the general problem of intra-granular fission gas diffusion during non-equilibrium trapping and resolution, which will be the subject of future work.

  2. Ternary fission of nuclei into comparable fragments

    SciTech Connect

    Karpeshin, F. F.

    2015-07-15

    The problem of nuclear fission into three comparable fragments is considered. A mechanism of true ternary fission is proposed. In contrast to sequential fission, where the three fragments arise upon two sequential events of binary fission, the mechanism in question relies on a scenario that originally involves fission into three fragments. This mechanism is driven by a hexadecapole deformation of the fissioning nucleus, in contrast to binary fission associated with quadrupole vibrations of the nuclear surface. The fragment-mass ratios are estimated. The dynamics of formation of collinear fragments and their subsequent motion in opposite directions is traced. The calculated probability of true ternary fission complies with observed values.

  3. Identification of failed fuel element

    DOEpatents

    Fryer, Richard M.; Matlock, Robert G.

    1976-06-22

    A passive fission product gas trap is provided in the upper portion of each fuel subassembly in a nuclear reactor. The gas trap consists of an inverted funnel of less diameter than the subassembly having a valve at the apex thereof. An actuating rod extends upwardly from the valve through the subassembly to a point where it can be contacted by the fuel handling mechanism for the reactor. Interrogation of the subassembly for the presence of fission products is accomplished by lowering the fuel handling machine onto the subassembly to press down on the actuating rod and open the valve.

  4. US/UK actinides experiment at the Dounreay PFR. 1: Fission products

    SciTech Connect

    Raman, S.; Murphy, B.D.

    1995-09-01

    The US and the United Kingdom have been engaged in a joint research program in which samples of higher actinides were irradiated in the 600-MW Dounreay Prototype Fast Reactor in Scotland. Analytical results using mass spectrometry and radiometry for actinides and fission products are now available for the samples in Fuel Pins 1 and 2 which were irradiated for 63 full-power days and for the samples in Fuel Pin 4 which were irradiated for 492 full-power days. Results from these three fuel pins are providing estimates of integral cross sections and fission yields.

  5. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Anderson, W.F.; Tellefson, D.R.; Shimazaki, T.T.

    1962-04-10

    A plate type fuel element which is particularly useful for organic cooled reactors is described. Generally, the fuel element comprises a plurality of fissionable fuel bearing plates held in spaced relationship by a frame in which the plates are slidably mounted in grooves. Clearance is provided in the grooves to allow the plates to expand laterally. The plates may be rigidly interconnected but are floatingly supported at their ends within the frame to allow for longi-tudinal expansion. Thus, this fuel element is able to withstand large temperature differentials without great structural stresses. (AEC)

  6. Thorium fueled reactor

    NASA Astrophysics Data System (ADS)

    Sipaun, S.

    2017-01-01

    Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

  7. TREATMENT OF FISSION PRODUCT WASTE

    DOEpatents

    Huff, J.B.

    1959-07-28

    A pyrogenic method of separating nuclear reactor waste solutions containing aluminum and fission products as buring petroleum coke in an underground retort, collecting the easily volatile gases resulting as the first fraction, he uminum chloride as the second fraction, permitting the coke bed to cool and ll contain all the longest lived radioactive fission products in greatly reduced volume.

  8. Membrane biology: fission behind BARs.

    PubMed

    Haucke, Volker

    2012-06-05

    Membrane bending is accomplished in part by amphipathic helix insertion into the bilayer and the assembly of BAR domain scaffolds preparing the membrane for fission. Two recent studies highlight the roles of amphipathic helices and BAR scaffolds in membrane fission and establish the structural basis of membrane bending by the N-BAR protein endophilin.

  9. Fission fragment driven neutron source

    DOEpatents

    Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

    1976-01-01

    Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

  10. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  11. Fission-suppressed fusion breeder on the thorium cycle and nonproliferation

    NASA Astrophysics Data System (ADS)

    Moir, R. W.

    2012-06-01

    Fusion reactors could be designed to breed fissile material while suppressing fissioning thereby enhancing safety. The produced fuel could be used to startup and makeup fuel for fission reactors. Each fusion reaction can produce typically 0.6 fissile atoms and release about 1.6 times the 14 MeV neutron's energy in the blanket in the fission-suppressed design. This production rate is 2660 kg/1000 MW of fusion power for a year. The revenues would be doubled from such a plant by selling fuel at a price of 60/g and electricity at 0.05/kWh for Q=Pfusion/Pinput=4. Fusion reactors could be designed to destroy fission wastes by transmutation and fissioning but this is not a natural use of fusion whereas it is a designed use of fission reactors. Fusion could supply makeup fuel to fission reactors that were dedicated to fissioning wastes with some of their neutrons. The design for safety and heat removal and other items is already accomplished with fission reactors. Whereas fusion reactors have geometry that compromises safety with a complex and thin wall separating the fusion zone from the blanket zone where wastes could be destroyed. Nonproliferation can be enhanced by mixing 233U with 238U. Also nonproliferation is enhanced in typical fission-suppressed designs by generating up to 0.05 232U atoms for each 233U atom produced from thorium, about twice the IAEA standards of "reduced protection" or "self protection." With 2.4% 232U, high explosive material is predicted to degrade owing to ionizing radiation after a little over 1/2 year and the heat rate is 77 W just after separation and climbs to over 600 W ten years later. The fissile material can be used to fuel most any fission reactor but is especially appropriate for molten salt reactors (MSR) also called liquid fluoride thorium reactors (LFTR) because of the molten fuel does not need hands on fabrication and handling.

  12. Fission-suppressed fusion breeder on the thorium cycle and nonproliferation

    SciTech Connect

    Moir, R. W.

    2012-06-19

    Fusion reactors could be designed to breed fissile material while suppressing fissioning thereby enhancing safety. The produced fuel could be used to startup and makeup fuel for fission reactors. Each fusion reaction can produce typically 0.6 fissile atoms and release about 1.6 times the 14 MeV neutron's energy in the blanket in the fission-suppressed design. This production rate is 2660 kg/1000 MW of fusion power for a year. The revenues would be doubled from such a plant by selling fuel at a price of 60/g and electricity at $0.05/kWh for Q=P{sub fusion}/P{sub input}=4. Fusion reactors could be designed to destroy fission wastes by transmutation and fissioning but this is not a natural use of fusion whereas it is a designed use of fission reactors. Fusion could supply makeup fuel to fission reactors that were dedicated to fissioning wastes with some of their neutrons. The design for safety and heat removal and other items is already accomplished with fission reactors. Whereas fusion reactors have geometry that compromises safety with a complex and thin wall separating the fusion zone from the blanket zone where wastes could be destroyed. Nonproliferation can be enhanced by mixing {sup 233}U with {sup 238}U. Also nonproliferation is enhanced in typical fission-suppressed designs by generating up to 0.05 {sup 232}U atoms for each {sup 233}U atom produced from thorium, about twice the IAEA standards of 'reduced protection' or 'self protection.' With 2.4%{sup 232}U, high explosive material is predicted to degrade owing to ionizing radiation after a little over 1/2 year and the heat rate is 77 W just after separation and climbs to over 600 W ten years later. The fissile material can be used to fuel most any fission reactor but is especially appropriate for molten salt reactors (MSR) also called liquid fluoride thorium reactors (LFTR) because of the molten fuel does not need hands on fabrication and handling.

  13. Nuclear fuel element

    DOEpatents

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  14. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Kesselring, K.A.; Seybolt, A.U.

    1958-12-01

    A reactor fuel element of the capillary tube type is described. The element consists of a thin walled tube, sealed at both ends, and having an interior coatlng of a fissionable material, such as uranium enriched in U-235. The tube wall is gas tight and is constructed of titanium, zirconium, or molybdenum.

  15. NEUTRONIC REACTOR FUEL PUMP

    DOEpatents

    Cobb, W.G.

    1959-06-01

    A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

  16. Early User Test: BISON for Fission Gas Release Modeling

    SciTech Connect

    Johns, Russell C.

    2014-01-28

    As a test to develop an understanding of the capabilities and input of the nuclear fuel performance code BISON, we modeled the fission gas release of the Oconee rod 15309 as there is benchmark data available for comparison. BISON has two fission gas models available, the Forsber-Massih model and the Simple Integrated Fission Gas Release and Swelling [sifgrs] model. Due to time constraints and availability of model parameters [we did not have all of the parameters for the sifgrs model already in hand] we only tested the Forsberg-Massih model. The Oconee rod test has been used in a number of benchmark studies including the integral assessment of FRAPCON 3.4. The Oconee Rod test is a 5-cycle test bundle with an average burnup of 50GWd/MTU. The FRAPCON assessment predicted that Oconee Rod15309 Test case would release 1.25% of all fission gas produced, while experimentally only 0.8% was released at EOL. Other studies using BISON with the sifgrs model have under predicted EOL fission gas release [FGR].

  17. Reference reactor module for NASA's lunar surface fission power system

    SciTech Connect

    Poston, David I; Kapernick, Richard J; Dixon, David D; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  18. Phase 1 Space Fission Propulsion System Testing and Development Progress

    NASA Technical Reports Server (NTRS)

    VanDyke, Melissa; Houts, Mike; Godfroy, Tom; Dickens, Ricky; Poston, David; Kapernick, Rick; Reid, Bob; Salvail, Pat; Ring, Peter; Schafer, Charles (Technical Monitor)

    2001-01-01

    Successful development of space fission systems requires an extensive program of affordable and realistic testing. In addition to tests related to design/development of the fission system, realistic testing of the actual flight unit must also be performed. If the system is designed to operate within established radiation damage and fuel burn up limits while simultaneously being designed to allow close simulation of heat from fission using resistance heaters, high confidence in fission system performance and lifetime can be attained through a series of non-nuclear tests. The Safe Affordable Fission Engine (SAFE) test series, whose ultimate goal is the demonstration of a 300 kW flight configuration system, has demonstrated that realistic testing can be performed using non-nuclear methods. This test series, carried out in collaboration with other NASA centers, other government agencies, industry, and universities, successfully completed a testing program with a 30 kWt core, Stirling engine, and ion engine configuration. Additionally, a 100 kWt core is in fabrication and appropriate test facilities are being reconfigured. This paper describes the current SAFE non-nuclear tests, which includes test article descriptions, test results and conclusions, and future test plans.

  19. Spontaneous fission properties of superheavy elements

    NASA Astrophysics Data System (ADS)

    Heßberger, F. P.

    2017-04-01

    Spontaneous fission properties of transuranium isotopes are reviewed. Specific emphasis was laid on brief historical overviews of theoretical descriptions and experimental determination of basic properties as spontaneous fission half-lives, fission barriers, or total kinetic energy release in fission. Experimental spontaneous fission half-lives are compared with the results of recent theoretical predictions. Hindrance factors for spontaneous fission of odd-mass nuclei are discussed in context with the configuration (spin, parity) of the fissioning states and the change in energy of single particle levels at deformation. Kinetic energy release and mass distributions are discussed in the context of different fission modes, as symmetric and asymmetric or fission from elongated or compact shapes of the nascent fission fragments. An overview of recent fission barrier calculations of superheavy elements on the basis of macroscopic-microscopic models or self-consistent calculations is given, and the results are compared for selected examples.

  20. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. )

    1989-01-01

    The news of the discovery of nuclear fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fifieth anniversary of its discovery by holding a topical meeting entitled, Fifty Years with Nuclear Fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent development in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicated a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two fully days of sessions (April 27 and 28) at the main site of the NIST in Gaithersburg, Maryland. The wide range of topics covered in this Volume 1 by this topical meeting included plenary invited, and contributed sessions entitled: Preclude to the First Chain Reaction -- 1932 to 1942; Early Fission Research -- Nuclear Structure and Spontaneous Fission; 50 Years of Fission, Science, and Technology; Nuclear Reactors, Secure Energy for the Future; Reactors 1; Fission Science 1; Safeguards and Space Applications; Fission Data; Nuclear Fission -- Its Various Aspects; Theory and Experiments in Support of Theory; Reactors and Safeguards; and General Research, Instrumentation, and By-Product. The individual papers have been cataloged separately.

  1. A review of selected aspects of the effect of water vapor on fission gas release from uranium oxycarbide

    SciTech Connect

    Myers, B.F.

    1994-04-01

    A selective review is presented of previous measurements and the analysis of experiments on the effect of water vapor on fission gas release from uranium oxycarbide. Evidence for the time-dependent composition of the uranium oxycarbide fuel; the diffusional release of fission gas; and the initial, rapid and limited release of stored fission gas is discussed. In regard to the initial, rapid release of fission gas, clear restrictions on mechanistic hypotheses can be deduced from the experimental data. However, more fundamental experiments may be required to establish the mechanism of the rapid release.

  2. Data summary report for fission product release test VI-5

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Travis, J.R.; Webster, C.S.; Collins, J.L. )

    1991-10-01

    Test VI-5, the fifth in a series of high-temperature fission product release tests in a vertical test apparatus, was conducted in a flowing mixture of hydrogen and helium. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium which had been irradiated to a burnup of {approximately}42 MWd/kg. Using a hot cell-mounted test apparatus, the fuel rod was heated in an induction furnace under simulated LWR accident conditions to two test temperatures, 2000 K for 20 min and then 2700 K for an additional 20 min. The released fission products were collected in three sequentially operated collection trains on components designed to measure fission product transport characteristics and facilitate sampling and analysis. The results from this test were compared with those obtained in previous tests in this series and with the CORSOR-M and ORNL diffusion release models for fission product release. 21 refs., 19 figs., 12 tabs.

  3. New results from the NSRR experiments with high burnup fuel

    SciTech Connect

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  4. Nuclear fuel element, and method of forming same

    SciTech Connect

    Taylor, I.N. Jr.; Magee, P.M.

    1990-11-20

    This patent describes improvement to the method of inhibiting an interaction between a metal alloy fissionable fuel for a nuclear reactor and a stainless steel container for the fuel which results in low temperature melting eutectic reaction products of components from the metal alloy fuel and stainless steel. The improvement comprises providing an expandable body of at least one alloying metal for the metal alloy fissionable fuel selected from the group consisting of zirconium, titanium, niobium and molybdenum at least about 2 mils thick interposed into a space between a metal alloy fuel. The fuel consists of metallic uranium and plutonium and the stainless steel container housing the metal alloy fuel therein.

  5. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  6. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    SciTech Connect

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  7. A compact multi-plate fission chamber for the simultaneous measurement of 233U capture and fission cross-sections

    NASA Astrophysics Data System (ADS)

    Bacak, M.; Berthoumieux, E.; Aiche, M.; Bélier, G.; Cardella, R.; Chatillon, A.; Diakaki, M.; Dupont, E.; Gunsing, F.; Heyse, J.; Kopecky, S.; Laurent, B.; Leeb, H.; Mathieu, L.; Schillebeeckx, P.; Taieb, J.; Vlachoudis, V.

    2017-09-01

    233U plays the essential role of fissile nucleus in the Th-U fuel cycle. A particularity of 233U is its small neutron capture cross-section which is about one order of magnitude lower than the fission cross-section on average. Therefore, the accuracy in the measurement of the 233U capture cross-section essentially relies on efficient capture-fission discrimination thus a combined setup of fission and γ-detectors is needed. At CERN n_TOF the Total Absorption Calorimeter (TAC) coupled with compact fission detectors is used. Previously used MicroMegas (MGAS) detectors showed significant γ-background issues above 100 eV coming from the copper mesh. A new measurement campaign of the 233U capture cross-section and alpha ratio is planned at the CERN n_TOF facility. For this measurement, a novel cylindrical multi ionization cell chamber was developed in order to provide a compact solution for 14 active targets read out by 8 anodes. Due to the high specific activity of 233U fast timing properties are required and achieved with the use of customized electronics and the very fast ionizing gas CF4 together with a high electric field strength. This paper describes the new fission chamber and the results of the first tests with neutrons at GELINA proving that it is suitable for the 233U measurement.

  8. Design and experimental investigation into fuel element melting during pulsed heating in the IGRIK

    SciTech Connect

    Levakov, B.G.; Andreev, V.V.; Vasilyev, A.P.

    1995-12-31

    Research has been performed on reactor fuel melting with pulsed input of energy in fuel elements up to 1.3 kj/g. The following were determined: energy input in fuel elements and energy input tempo; fission number distribution by the radius of the fuel element; the temperature of fuel and ampoule walls; and displacement of fuel boundaries.

  9. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  10. Dynamical Aspects of Nuclear Fission

    NASA Astrophysics Data System (ADS)

    Kliman, J.; Itkis, M. G.; Gmuca, Š.

    2008-11-01

    Fission dynamics. Dependence of scission-neutron yield on light-fragment mass for [symbol]=1/2 [et al.]. Dynamics of capture quasifission and fusion-fission competition / L. Stuttgé ... [et al.] -- Fission-fission. The processes of fusion-fission and quasi-fission of superheavy nuclei / M. G. Itkis ... [et al.]. Fission and quasifission in the reactions [symbol]Ca+[symbol]Pb and [symbol]Ni+[symbol]W / G. N. Knyazheva ... [et al.]. Mass-energy characteristics of reactions [symbol]Fe+[symbol][symbol][symbol]266Hs and [symbol]Mg+[symbol]Cm[symbol][symbol]Hs at Coulomb barrier / L. Krupa ... [et al.]. Fusion of heavy ions at extreme sub-barrier energies / Ş. Mişicu and H. Esbensen. Fusion and fission dynamics of heavy nuclear system / V. Zagrebaev and W. Greiner. Time-dependent potential energy for fusion and fission processes / A. V. Karpov ... [et al.] -- Superheavy elements. Advances in the understanding of structure and production mechanisms for superheavy elements / W. Greiner and V. Zagrebaev. Fission barriers of heaviest nuclei / A. Sobiczewski ... [et al.]. Possibility of synthesizing doubly magic superheavy nuclei / Y Aritomo ... [et al.]. Synthesis of superheavy nuclei in [symbol]Ca-induced reactions / V. K. Utyonkov ... [et al.] -- Fragmentation. Production of neutron-rich nuclei in the nucleus-nucleus collisions around the Fermi energy / M. Veselský. Signals of enlarged core in [symbol]Al / Y. G. Ma ... [et al.] -- Exotic modes. New insight into the fission process from experiments with relativistic heavy-ion beams / K.-H. Schmidt ... [et al.]. New results for the intensity of bimodal fission in binary and ternary spontaneous fission of [symbol]Cf / C. Goodin ... [et al.]. Rare fission modes: study of multi-cluster decays of actinide nuclei / D. V. Kamanin ... [et al.]. Energy distribution of ternary [symbol]-particles in [symbol]Cf(sf) / M. Mutterer ... [et al.]. Preliminary results of experiment aimed at searching for collinear cluster tripartition of

  11. Radiological Aspects of Deep-Burn Fusion-Fission Hybrid Waste in a Repository

    SciTech Connect

    Shaw, H F; Blink, J A; Farmer, J C; Karmer, K J; Latkowski, J F; Zhao, P

    2008-11-25

    The quantity, radioactivity, and isotopic characteristics of the spent fission fuel from a hybrid fusion-fission system capable of extremely high burnups are described. The waste generally has higher activity per unit mass of heavy metal, but much lower activity per unit energy generated. The very long-term radioactivity is dominated by fission products. Simple scaling calculations suggest that the dose from a repository containing such waste would be dominated by {sup 129}I, {sup 135}Cs, and {sup 242}Pu. Use of such a system for generating energy would greatly reduce the need for repository capacity.

  12. Fission thrust sail as booster for high Δv fusion based propulsion

    NASA Astrophysics Data System (ADS)

    Ceyssens, Frederik; Wouters, Kristof; Driesen, Maarten

    2015-12-01

    The fission thrust sail as booster for nuclear fusion-based rocket propulsion for future starships is introduced and studied. First order calculations are used together with Monte Carlo simulations to assess system performance. If a D-D fusion rocket such as e.g. considered in Project Icarus has relatively low efficiency (~30%) in converting fusion fuel to a directed exhaust, adding a fission sail is shown to be beneficial for the obtainable delta-v. In addition, this type of fission-fusion hybrid propulsion has the potential to improve acceleration and act as a micrometeorite shield.

  13. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. )

    1989-01-01

    The news of the discovery of nucler fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fiftieth anniversary of its discovery by holding a topical meeting entitled, Fifty years with nuclear fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent developments in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicating a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two full days of sessions (April 27 and 28) at the main sites of the NIST in Gaithersburg, Maryland. The wide range of topics covered by Volume 2 of this topical meeting included plenary invited, and contributed sessions entitled, Nuclear fission -- a prospective; reactors II; fission science II; medical and industrial applications by by-products; reactors and safeguards; general research, instrumentation, and by-products; and fission data, astrophysics, and space applications. The individual papers have been cataloged separately.

  14. Neutron-induced fission cross section of 233Pa between 1.0 and 3.0 MeV.

    PubMed

    Tovesson, F; Hambsch, F J; Oberstedt, A; Fogelberg, B; Ramström, E; Oberstedt, S

    2002-02-11

    The energy dependent neutron-induced fission cross section of 233Pa has for the first time been measured directly with monoenergetic neutrons. This nuclide is an important intermediary in a thorium based fuel cycle, and its fission cross section is a key parameter in the modeling of future advanced fuel and reactor concepts. A first experiment resulted in four cross section values between 1.0 and 3.0 MeV, establishing a fission threshold in excess of 1 MeV. Significant discrepancies were found with a previous indirect experimental determination and with model estimates.

  15. Spallation-induced fission reactions

    NASA Astrophysics Data System (ADS)

    Benlliure, J.; Rodríguez-Sánchez, J. L.

    2017-03-01

    During the last decade spallation-induced fission reactions have received particular attention because of their impact in the design of spallation-neutron sources or radioactive beam facilities, but also in the understanding of the fission process at high excitation energy. In this paper, we review the main progress brought by modern experimental techniques, in particular those based in the inverse kinematic, as well as the achievements in modelling these reactions. We will also address future possibilities for improving the investigation of fission dynamics.

  16. Fission at intermediate nucleon energies

    NASA Astrophysics Data System (ADS)

    Lo Meo, S.; Mancusi, D.; Massimi, C.; Vannini, G.; Ventura, A.

    2014-07-01

    In the present work Monte Carlo calculations of fission of actinides and pre- actinides induced by protons and neutrons in the energy range from 100 MeV to 1 GeV are carried out by means of a recent version of the Liège Intranuclear Cascade Model, INCL++, coupled with different evaporation-fission codes, in particular GEMINI++ and ABLA07. Fission model parameters are adjusted on experimental (p, f) cross sections and used to predict (n, f) cross sections, in order to provide a theoretical support to the campaign of neutron cross section measurements at the n_TOF facility at CERN.

  17. CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT ...

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

    CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT BREAKS IN CLADDING OF FUEL ELEMENTS. COUNT-RATE METER IN TOP PANEL INDICATES AMOUNT OF RADIOACTIVITY. LOWER PANELS SUPPLY POWER AND AMPLIFICATION OF SIGNALS GENERATED BY SCINTILLATION COUNTER/PHOTOMULTIPLIER TUBE COMBINATION IN RESPONSE TO RADIOACTIVITY IN A SAMPLE OF THE COOLING WATER. INL NEGATIVE NO. 56-771. Jack L. Anderson, Photographer, 3/15/1956. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  18. Mini-fission fusion explosive devices (mini-nukes) for nuclear pulse propulsion

    NASA Astrophysics Data System (ADS)

    Winterberg, F.

    2005-11-01

    Nuclear pulse propulsion demands low-yield nuclear explosive devices. Because the critical mass of a fission explosive is rather large, this leads to extravagant fission devices with a very low fuel burn-up. For non-fission ignited pure fusion microexplosions the problem is the large ignition apparatus (laser, particle beam, etc.). Fission ignited large fusion explosive devices are for obvious reasons even less desirable. A third category (mini-nukes) are devices where the critical mass of the fission explosive is substantially reduced by its coupling to a DT fusion reaction, with the DT fusion neutrons increasing the fission rate. Whereas in pure fission devices a reduction of the critical mass is achieved by the implosive compression of the fissile core with a chemical high explosive, in the third category the implosion must at the same time heat the DT surrounding the fissile core to a temperature of ⩾107K, at which enough fusion neutrons are generated to increase the fission rate which in turn further increases the temperature and fusion neutron production rate. As has been shown by the author many years ago, such mini-nukes lead to astonishingly small critical masses. In their application to nuclear pulse propulsion the combustion products from the chemical high explosive are further heated by the neutrons and are becoming part of the propellant.

  19. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-16

    A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

  20. Swelling-resistant nuclear fuel

    SciTech Connect

    Arsenlis, Athanasios; Satcher, Jr., Joe; Kucheyev, Sergei O

    2011-12-27

    A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

  1. Theoretical investigation of fission fragment kinetic energy distributions in the symmetric mass region for 233U(nth,f)

    NASA Astrophysics Data System (ADS)

    Chebboubi, Abdelaziz; Serot, Olivier; Kessedjian, Grégoire; Litaize, Olivier; Blanc, Aurelien; Bernard, David; Faust, Herbert; Julien-Laferrière, Sylvain; Köster, Ulli; Letourneau, Alain; Materna, Thomas; Méplan, Olivier; Mutti, Paolo; Rapala, Michal; Sage, Christophe

    2017-09-01

    Fission yields are essential for nuclear reactor studies (decay heat, fuel inventory…) and constitute also one of the main observables needed to improve our understanding of the fission process. The symmetric mass region is of particular interest due to various intriguing properties of the fission fragments already reported in the literature : inversion of the nuclear charge polarization, large width of the fission fragment kinetic energy distribution, strong change of the prompt neutron multiplicity, etc. Recently, measurements of fission yields and kinetic energy distributions in the symmetric mass region were achieved at the LOHENGRIN mass spectrometer of the Institut Laue-Langevin (ILL). This experimental work is challenging due to the low counting rate and the appearance of contaminant masses, leading to pronounced components in the fission fragment kinetic energy distribution. Despite removing the undesirable contributions, the fission fragment kinetic energy distributions still show two components, indicating that the fission process could be modal. To go further and better characterize these components a comparison between our experimental data and Monte Carlo calculations (FIFRELIN code) simulating the de-excitation of the fission fragments for different fission channels will be presented and discussed.

  2. The binary fission origin of the moon

    NASA Technical Reports Server (NTRS)

    Binder, Alan B.

    1986-01-01

    The major arguments for and against the binary fission model of lunar origin are reviewed. Unresolved problems include: (1) how the protoearth acquired sufficient angular velocity to fission, and (2) how the earth-moon system lost its excess angular momentum after fission. Despite these uncertainties, the compositional similarities between the earth's mantle and the bulk moon suggest that the fission model is worth considering. The proposed sequence of events in the formation of the moon by binary fission is given.

  3. Fission gas bubble identification using MATLAB's image processing toolbox

    SciTech Connect

    Collette, R.; King, J.; Keiser, Jr., D.; Miller, B.; Madden, J.; Schulthess, J.

    2016-06-08

    Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. In addition, this study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding proved to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods.

  4. Fission gas bubble identification using MATLAB's image processing toolbox

    DOE PAGES

    Collette, R.; King, J.; Keiser, Jr., D.; ...

    2016-06-08

    Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. In addition, this study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding provedmore » to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods.« less

  5. 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.

  6. Fission gas bubble identification using MATLAB's image processing toolbox

    SciTech Connect

    Collette, R.; King, J.; Keiser, Jr., D.; Miller, B.; Madden, J.; Schulthess, J.

    2016-06-08

    Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. In addition, this study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding proved to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods.

  7. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    ERIC Educational Resources Information Center

    Settle, Frank A.

    2009-01-01

    The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

  8. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    ERIC Educational Resources Information Center

    Settle, Frank A.

    2009-01-01

    The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

  9. Fission of rotating fermium isotopes

    NASA Astrophysics Data System (ADS)

    Baran, A.; Staszczak, A.

    2014-05-01

    In this paper we discuss the process of fission of even fermium isotopes, on the basis of their rotational states. The nuclear intrinsic vorticity and its coupling to the global rotation of the nucleus are used to simulate the interaction between the rotational motion and the pairing field, and lead to pairing quenching in the case of higher angular momentum states. The rotation leads to a decreasing of the fission barrier heights. The ingredients of the model—ground state fission barriers, pairing correlation energies and the cranking moments of inertia—are obtained within the self-consistent Hartree-Fock-Bogoliubov framework using the Skyrme \\text{Sk}{{\\text{M}}^{*}} energy density functional. Fission barriers and half-lives are estimated for spins I up to I = 16ℏ.

  10. Ternary Fission of CF Isotopes

    NASA Astrophysics Data System (ADS)

    Vermote, S.; Wagemans, C.; Serot, O.; Soldner, T.; Geltenbort, P.; Almahamid, I.; Lukens, W.; Floyd, J.

    2008-04-01

    During the last years, different Cm and Cf isotopes have been studied by our research group in the frame of a systematic investigation of gas emission characteristics in ternary fission. In this paper we report on the energy distribution and the emission probability of 3H, 4He and 6He particles emitted in neutron induced ternary fission of 249Cf and 251Cf. Both measurements were performed at the high flux reactor of the Institute Laue-Langevin (Grenoble, France), using suited ΔE-E telescope detectors, consisting of well-calibrated silicon surface barrier detectors. In this way, the available database can be expanded with new results for Z=98 isotopes, for which the information on neutron induced ternary fission is almost nonexistent. These measurements are important for the systematic investigation of gas emission characteristics in ternary fission.

  11. The Microscopic Theory of Fission

    SciTech Connect

    Younes, W; Gogny, D

    2009-06-09

    Fission-fragment properties have been calculated for thermal neutron-induced fission on a {sup 239}Pu target, using constrained Hartree-Fock-Bogoliubov calculations with a finite-range effective interaction. A quantitative criterion based on the interaction energy between the nascent fragments is introduced to define the scission configurations. The validity of this criterion is benchmarked against experimental measurements of the kinetic energies and of multiplicities of neutrons emitted by the fragments.

  12. Bulk and surface controlled diffusion of fission gas atoms

    SciTech Connect

    Andersson, Anders D.

    2012-08-09

    Fission gas retention and release impact nuclear fuel performance by, e.g., causing fuel swelling leading to mechanical interaction with the clad, increasing the plenum pressure and reducing the gap thermal conductivity. All of these processes are important to understand in order to optimize operating conditions of nuclear reactors and to simulate accident scenarios. Most fission gases have low solubility in the fuel matrix, which is especially pronounced for large fission gas atoms such as Xe and Kr, and as a result there is a significant driving force for segregation of gas atoms to extended defects such as grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. Several empirical or semi-empirical models have been developed for fission gas release in nuclear fuels, e.g. [1-6]. One of the most commonly used models in fuel performance codes was published by Massih and Forsberg [3,4,6]. This model is similar to the early Booth model [1] in that it applies an equivalent sphere to separate bulk UO{sub 2} from grain boundaries represented by the sphere circumference. Compared to the Booth model, it also captures trapping at grain boundaries, fission gas resolution and it describes release from the boundary by applying timedependent boundary conditions to the circumference. In this work we focus on the step where fission gas atoms diffuse from the grain interior to the grain boundaries. The original Massih-Forsberg model describes this process by applying an effective diffusivity divided into three temperature regimes. In this report we present results from density functional theory calculations (DFT) that are relevant for the high (D{sub 3}) and intermediate (D{sub 2}) temperature diffusivities of fission gases. The results are validated by making a quantitative comparison to Turnbull's [8-10] and Matzke's data [12]. For the intrinsic or high temperature regime we report activation energies for both Xe and Kr diffusion in UO

  13. Nuclear Design of the SAFE-400a Space Fission Reactor

    SciTech Connect

    Poston, David I.

    2002-07-01

    Ambitious solar system exploration missions in the near future will require robust power sources in the range of 10 to 200 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The Heat pipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed - which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Safe Affordable Fission Engine (SAFE) is an HPS reactor designed for producing electricity in space. The SAFE-400a is a 400- kWt reactor that has been designed to couple with a 100-kWe Brayton power system. The SAFE-400a contains 127 identical molybdenum (Mo) modules. A Mo/sodium heat pipe is at the center of each module, surrounded by three Mo tubes that each contain a rhenium-clad uranium-nitride fuel pin. Fission energy is conducted from the fuel pins to the heat pipes, which then carry the heat to a heat pipe-to-gas heat exchanger. This paper describes the design and analysis of the current SAFE-400a reactor design. (author)

  14. Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

    SciTech Connect

    Poston, David I.

    2002-07-01

    The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heat pipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed - which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heat pipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heat pipes. Fission energy is conducted from the fuel pins to the heat pipes, which then carry the heat to the Stirling engine. This paper describes conceptual design and nuclear performance the HOMER-15 reactor. (author)

  15. Design and analysis of the SAFE-400 space fission reactor

    NASA Astrophysics Data System (ADS)

    Poston, David I.; Kapernick, Richard J.; Guffee, Ray M.

    2002-01-01

    Ambitious solar system exploration missions in the near future will require robust power sources in the range of 10 to 200 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The Heatpipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed-which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Safe Affordable Fission Engine (SAFE) is an HPS reactor designed for producing electricity in space. The SAFE-400 is a 400-kWt reactor that has been designed to couple with a 100-kWe Brayton power system. The SAFE-400 contains 127 identical molybdenum (Mo) modules. A Mo/sodium heatpipe is at the center of each module, surrounded by three Mo tubes that each contain a rhenium-clad uranium-nitride fuel pin. Fission energy is conducted from the fuel pins to the heatpipes, which then carry the heat to a heatpipe-to-gas heat exchanger. This paper describes the design and analysis of the current SAFE-400 reactor design. .

  16. Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    SciTech Connect

    Blaise Collin

    2014-09-01

    This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive

  17. Nuclear fission and the transuranium elements

    SciTech Connect

    Seaborg, G.T.

    1989-02-01

    Many of the transuranium elements are produced and isolated in large quantities through the use of neutrons furnished by nuclear fission reactions: plutonium (atomic number 94) in ton quantities; neptunium (93), americium (95), and curium (96) in kilogram quantities; berkelium (97) in 100 milligram quantities; californium (98) in gram quantities; and einsteinium (99) in milligram quantities. Transuranium isotopes have found many practical applications---as nuclear fuel for the large-scale generation of electricity, as compact, long-lived power sources for use in space exploration, as means for diagnosis and treatment in the medical area, and as tools in numerous industrial processes. Of particular interest is the unusual chemistry and impact of these heaviest elements on the periodic table. This account will feature these aspects. 9 refs., 5 figs.

  18. Lasers from fission. [nuclear pumping feasibility experiments

    NASA Technical Reports Server (NTRS)

    Schneider, R. T.; Thom, K.; Helmick, H. H.

    1975-01-01

    The feasibility of the nuclear pumping of lasers was demonstrated in three experiments conducted independently at three different laboratories. In this context nuclear pumping of lasers is understood to be the excitation of a laser by the kinetic energy of the fission fragments only. A description is given of research concerned with the use of nuclear energy for the excitation of gas lasers. Experimental work was supplemented by theoretical research. Attention is given to a nuclear pumped He-Xe laser, a nuclear pumped CO laser, and a neon-nitrogen laser pumped by alpha particles. Studies involving uranium hexafluoride admixture to laser media are discussed along with research on uranium hexafluoride-fueled reactors.

  19. Lasers from fission. [nuclear pumping feasibility experiments

    NASA Technical Reports Server (NTRS)

    Schneider, R. T.; Thom, K.; Helmick, H. H.

    1975-01-01

    The feasibility of the nuclear pumping of lasers was demonstrated in three experiments conducted independently at three different laboratories. In this context nuclear pumping of lasers is understood to be the excitation of a laser by the kinetic energy of the fission fragments only. A description is given of research concerned with the use of nuclear energy for the excitation of gas lasers. Experimental work was supplemented by theoretical research. Attention is given to a nuclear pumped He-Xe laser, a nuclear pumped CO laser, and a neon-nitrogen laser pumped by alpha particles. Studies involving uranium hexafluoride admixture to laser media are discussed along with research on uranium hexafluoride-fueled reactors.

  20. Fuel System Compatibility Issues for Prometheus-1

    SciTech Connect

    DC Noe; KB Gibbard; MH Krohn

    2006-01-20

    Compatibility issues for the Prometheus-1 fuel system have been reviewed based upon the selection of UO{sub 2} as the reference fuel material. In particular, the potential for limiting effects due to fuel- or fission product-component (cladding, liner, spring, etc) chemical interactions and clad-liner interactions have been evaluated. For UO{sub 2}-based fuels, fuel-component interactions are not expected to significantly limit performance. However, based upon the selection of component materials, there is a potential for degradation due to fission products. In particular, a chemical liner may be necessary for niobium, tantalum, zirconium, or silicon carbide-based systems. Multiple choices exist for the configuration of a chemical liner within the cladding; there is no clear solution that eliminates all concerns over the mechanical performance of a clad/liner system. A series of tests to evaluate the performance of candidate materials in contact with real and simulated fission products is outlined.

  1. Nuclear reactor fuel element having improved heat transfer

    DOEpatents

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  2. Fission Surface Power for the Exploration and Colonization of Mars

    NASA Technical Reports Server (NTRS)

    Houts, Mike; Porter, Ron; Gaddis, Steve; Van Dyke, Melissa; Martin, Jim; Godfroy, Tom; Bragg-Sitton, Shannon; Garber, Anne; Pearson, Boise

    2006-01-01

    The colonization of Mars will require abundant energy. One potential energy source is nuclear fission. Terrestrial fission systems are highly developed and have the demonstrated ability to safely produce tremendous amounts of energy. In space, fission systems not only have the potential to safely generate tremendous amounts of energy, but could also potentially be used on missions where alternatives are not practical. Programmatic risks such as cost and schedule are potential concerns with fission surface power (FSP) systems. To be mission enabling, FSP systems must be affordable and programmatic risk must be kept acceptably low to avoid jeopardizing exploration efforts that may rely on FSP. Initial FSP systems on Mars could be "workhorse" units sized to enable the establishment of a Mars base and the early growth of a colony. These systems could be nearly identical to FSP systems used on the moon. The systems could be designed to be safe, reliable, and have low development and recurring costs. Systems could also be designed to fit on relatively small landers. One potential option for an early Mars FSP system would be a 100 kWt class, NaK cooled system analogous to space reactors developed and flown under the U.S. "SNAP" program or those developed and flown by the former Soviet Union ("BUK" reactor). The systems could use highly developed fuel and materials. Water and Martian soil could be used to provide shielding. A modern, high-efficiency power conversion subsystem could be used to reduce required reactor thermal power. This, in turn, would reduce fuel burnup and radiation damage .effects by reducing "nuclear" fuels and materials development costs. A realistic, non-nuclear heated and fully integrated technology demonstration unit (TDU) could be used to reduce cost and programmatic uncertainties prior to initiating a flight program.

  3. Fission Surface Power for the Exploration and Colonization of Mars

    NASA Technical Reports Server (NTRS)

    Houts, Mike; Porter, Ron; Gaddis, Steve; Van Dyke, Melissa; Martin, Jim; Godfroy, Tom; Bragg-Sitton, Shannon; Garber, Anne; Pearson, Boise

    2006-01-01

    The colonization of Mars will require abundant energy. One potential energy source is nuclear fission. Terrestrial fission systems are highly developed and have the demonstrated ability to safely produce tremendous amounts of energy. In space, fission systems not only have the potential to safely generate tremendous amounts of energy, but could also potentially be used on missions where alternatives are not practical. Programmatic risks such as cost and schedule are potential concerns with fission surface power (FSP) systems. To be mission enabling, FSP systems must be affordable and programmatic risk must be kept acceptably low to avoid jeopardizing exploration efforts that may rely on FSP. Initial FSP systems on Mars could be "workhorse" units sized to enable the establishment of a Mars base and the early growth of a colony. These systems could be nearly identical to FSP systems used on the moon. The systems could be designed to be safe, reliable, and have low development and recurring costs. Systems could also be designed to fit on relatively small landers. One potential option for an early Mars FSP system would be a 100 kWt class, NaK cooled system analogous to space reactors developed and flown under the U.S. "SNAP" program or those developed and flown by the former Soviet Union ("BUK" reactor). The systems could use highly developed fuel and materials. Water and Martian soil could be used to provide shielding. A modern, high-efficiency power conversion subsystem could be used to reduce required reactor thermal power. This, in turn, would reduce fuel burnup and radiation damage .effects by reducing "nuclear" fuels and materials development costs. A realistic, non-nuclear heated and fully integrated technology demonstration unit (TDU) could be used to reduce cost and programmatic uncertainties prior to initiating a flight program.

  4. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    DOE PAGES

    Collette, R.; King, J.; Buesch, C.; ...

    2016-04-01

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

  5. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    NASA Astrophysics Data System (ADS)

    Collette, R.; King, J.; Buesch, C.; Keiser, D. D.; Williams, W.; Miller, B. D.; Schulthess, J.

    2016-07-01

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. The results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.

  6. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    SciTech Connect

    Collette, R.; King, J.; Buesch, C.; Keiser, Jr., D. D.; Williams, W.; Miller, B. D.; Schulthess, J.

    2016-04-01

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.

  7. Recent MELCOR and VICTORIA Fission Product Research at the NRC

    SciTech Connect

    Bixler, N.E.; Cole, R.K.; Gauntt, R.O.; Schaperow, J.H.; Young, M.F.

    1999-01-21

    thermochemistry was also improved, and results in better prediction of vaporization of uranium from fuel, which can react with released fission products to affect their volatility. This model also improves the prediction of fission product release rates from fuel. Finally, recent comparisons of MELCOR and VICTORIA with International Standard Problem 40 (STORM) data are presented. These comparisons focus on predicted therrnophoretic deposition, which is the dominant deposition mechanism. Sensitivity studies were performed with the codes to examine experimental and modeling uncertainties.

  8. Neutron Emission in Fission And Quasi-Fission of Hs

    SciTech Connect

    Itkis, I. M.; Itkis, M. G.; Knyazheva, G. N.; Kozulin, E. M.; Krupa, L.; Hanappe, F.; Dorvaux, O.; Stuttge, L.

    2010-04-30

    Mass and energy distributions of fission-like fragments obtained in the reactions {sup 26}Mg+{sup 248}Cm, {sup 36}S+{sup 238}U and {sup 58}Fe+{sup 208}Pb leading to the formation of {sup 266,274}Hs are reported. From the analysis of TKE distributions for symmetric fragment it was found that at energies below the Coulomb barrier the bimodal fission of {sup 274}Hs, formed in the reaction {sup 26}Mg+{sup 248}Cm, is observed, while in the reaction {sup 36}S+{sup 238}U at these energies the main part of the symmetric fragments arises from the quasi-fission process. At energies above the Coulomb barrier the fusion-fission is a main process leading to the formation of symmetric fragment for the both reactions. In the case of {sup 58}Fe+{sup 208}Pb reaction the quasi-fission process is the main reaction mechanism at all measured energies. The pre- and post-scission neutron multiplicities as a function of the fragment mass have been obtained for all studied reactions.

  9. Validation of the fission yield and decay data libraries with the 10 s-delayed 235 U fission γ-ray energy spectrum

    NASA Astrophysics Data System (ADS)

    Mendoza, E.; Álvarez-Velarde, F.; Bécares, V.; Cano-Ott, D.; González-Romero, E.; Martínez, T.; Villamarín, D.

    2017-10-01

    We have measured with a LaCl3 detector the γ-ray spectrum emitted by a 235 U enriched UO2 fuel rod 10 s after being irradiated with thermal neutrons. The experimental results are compared with simulations performed with the fission product yield and radioactive decay data libraries present in the most recent releases of ENDF/B, JEFF and JENDL.

  10. Reference Reactor Module for the Affordable Fission Surface Power System

    NASA Astrophysics Data System (ADS)

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO2-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important ``affordability'' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  11. Reference Reactor Module for the Affordable Fission Surface Power System

    SciTech Connect

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-21

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO{sub 2}-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important 'affordability' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  12. Energy production using fission fragment rockets

    SciTech Connect

    Chapline, G.; Matsuda, Y.

    1991-08-01

    Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: Approximately twice as efficient if one can directly convert the fission fragment energy into electricity; by reducing the buildup of a fission fragment inventory in the reactor one could avoid a Chernobyl type disaster; and collecting the fission fragments outside the reactor could simplify the waste disposal problem. 6 refs., 4 figs., 2 tabs.

  13. Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Albright, Dennis; Butler, Carey; West, Nicole; Cole, John W. (Technical Monitor)

    2002-01-01

    Institute for Scientific Research, Inc. (ISR) research program consist of: 1.Study core physics by adapting existing codes: MCNP4C - Monte Carlo code; COMBINE/VENTURE - diffusion theory; SCALE4 - Monte Carlo, with many utility codes. 2. Determine feasibility and study major design parameters: fuel selection, temperature and reflector sizing. 3. Study reactor kinetics: develop QCALC1 to model point kinetics; study dynamic behavior of the power release.

  14. Accelerator-Driven Thorium-Cycle Fission:. Green Nuclear Power for the New Millennium

    NASA Astrophysics Data System (ADS)

    McIntyre, Peter; Sattarov, Akhdiyor

    2011-03-01

    In thorium-cycle fission, fast neutrons are used to transmute thorium to fissionable 233U and then stimulate fission. In accelerator-driven thorium-cycle fission (ADTC) the fast neutrons are produced by injecting a symmetric pattern of 7 energetic proton beams into a Pb spallation zone in the core. The fast neutrons are adiabatically moderated by the Pb so that they capture efficiently on 232Th, and fission heat is transferred via a convective Pb column above the core. The 7 proton beams are generated by a flux-coupled stack of isochronous cyclotrons. ADTC offers a green solution to the Earth's energy needs: the core operates as a sub-critical pile and cannot melt down; it eats its own long-lived fission products; a GW ADTC core can operate with uniform power density for a 7-year fuel cycle without shuffling fuel pins, and there are sufficient thorium reserves to run man's energy needs for the next 2000 years.

  15. Fission fragment charge and mass distributions in 239Pu(n, f ) in the adiabatic nuclear energy density functional theory

    SciTech Connect

    Regnier, D.; Dubray, N.; Schunck, N.; Verriere, M.

    2016-05-13

    Here, accurate knowledge of fission fragment yields is an essential ingredient of numerous applications ranging from the formation of elements in the r process to fuel cycle optimization for nuclear energy. The need for a predictive theory applicable where no data are available, together with the variety of potential applications, is an incentive to develop a fully microscopic approach to fission dynamics.

  16. Fission fragment charge and mass distributions in 239Pu(n, f ) in the adiabatic nuclear energy density functional theory

    SciTech Connect

    Regnier, D.; Dubray, N.; Schunck, N.; Verriere, M.

    2016-05-13

    Here, accurate knowledge of fission fragment yields is an essential ingredient of numerous applications ranging from the formation of elements in the r process to fuel cycle optimization for nuclear energy. The need for a predictive theory applicable where no data are available, together with the variety of potential applications, is an incentive to develop a fully microscopic approach to fission dynamics.

  17. Compact fission counter for DANCE

    SciTech Connect

    Wu, C Y; Chyzh, A; Kwan, E; Henderson, R; Gostic, J; Carter, D; Bredeweg, T; Couture, A; Jandel, M; Ullmann, J

    2010-11-06

    The Detector for Advanced Neutron Capture Experiments (DANCE) consists of 160 BF{sub 2} crystals with equal solid-angle coverage. DANCE is a 4{pi} {gamma}-ray calorimeter and designed to study the neutron-capture reactions on small quantities of radioactive and rare stable nuclei. These reactions are important for the radiochemistry applications and modeling the element production in stars. The recognition of capture event is made by the summed {gamma}-ray energy which is equivalent of the reaction Q-value and unique for a given capture reaction. For a selective group of actinides, where the neutron-induced fission reaction competes favorably with the neutron capture reaction, additional signature is needed to distinguish between fission and capture {gamma} rays for the DANCE measurement. This can be accomplished by introducing a detector system to tag fission fragments and thus establish a unique signature for the fission event. Once this system is implemented, one has the opportunity to study not only the capture but also fission reactions. A parallel-plate avalanche counter (PPAC) has many advantages for the detection of heavy charged particles such as fission fragments. These include fast timing, resistance to radiation damage, and tolerance of high counting rate. A PPAC also can be tuned to be insensitive to {alpha} particles, which is important for experiments with {alpha}-emitting actinides. Therefore, a PPAC is an ideal detector for experiments requiring a fast and clean trigger for fission. A PPAC with an ingenious design was fabricated in 2006 by integrating amplifiers into the target assembly. However, this counter was proved to be unsuitable for this application because of issues related to the stability of amplifiers and the ability to separate fission fragments from {alpha}'s. Therefore, a new design is needed. A LLNL proposal to develop a new PPAC for DANCE was funded by NA22 in FY09. The design goal is to minimize the mass for the proposed counter

  18. Thorium-uranium fission radiography

    NASA Technical Reports Server (NTRS)

    Haines, E. L.; Weiss, J. R.; Burnett, D. S.; Woolum, D. S.

    1976-01-01

    Results are described for studies designed to develop routine methods for in-situ measurement of the abundance of Th and U on a microscale in heterogeneous samples, especially rocks, using the secondary high-energy neutron flux developed when the 650 MeV proton beam of an accelerator is stopped in a 42 x 42 cm diam Cu cylinder. Irradiations were performed at three different locations in a rabbit tube in the beam stop area, and thick metal foils of Bi, Th, and natural U as well as polished silicate glasses of known U and Th contents were used as targets and were placed in contact with mica which served as a fission track detector. In many cases both bare and Cd-covered detectors were exposed. The exposed mica samples were etched in 48% HF and the fission tracks counted by conventional transmitted light microscopy. Relative fission cross sections are examined, along with absolute Th track production rates, interaction tracks, and a comparison of measured and calculated fission rates. The practicality of fast neutron radiography revealed by experiments to data is discussed primarily for Th/U measurements, and mixtures of other fissionable nuclei are briefly considered.

  19. Thorium-uranium fission radiography

    NASA Technical Reports Server (NTRS)

    Haines, E. L.; Weiss, J. R.; Burnett, D. S.; Woolum, D. S.

    1976-01-01

    Results are described for studies designed to develop routine methods for in-situ measurement of the abundance of Th and U on a microscale in heterogeneous samples, especially rocks, using the secondary high-energy neutron flux developed when the 650 MeV proton beam of an accelerator is stopped in a 42 x 42 cm diam Cu cylinder. Irradiations were performed at three different locations in a rabbit tube in the beam stop area, and thick metal foils of Bi, Th, and natural U as well as polished silicate glasses of known U and Th contents were used as targets and were placed in contact with mica which served as a fission track detector. In many cases both bare and Cd-covered detectors were exposed. The exposed mica samples were etched in 48% HF and the fission tracks counted by conventional transmitted light microscopy. Relative fission cross sections are examined, along with absolute Th track production rates, interaction tracks, and a comparison of measured and calculated fission rates. The practicality of fast neutron radiography revealed by experiments to data is discussed primarily for Th/U measurements, and mixtures of other fissionable nuclei are briefly considered.

  20. Fission Rate Ratios of FCA-IX Assemblies as Integral Experiment for Assessment of TRU's Fission Cross Sections

    NASA Astrophysics Data System (ADS)

    Fukushima, Masahiro; Tsujimoto, Kazufumi; Okajima, Shigeaki

    2016-03-01

    At the fast critical assembly (FCA) of JAEA, central fission rate ratios for TRU such as 237Np, 238Pu, 239Pu, 242Pu, 241Am, 243Am, and 244Cm were measured in the seven uraniumfueled assemblies (FCA-IX assemblies) with systematically changed neutron spectra. The FCA-IX assemblies were constructed with simplicity both in geometry and composition. By virtue of these FCA-IX assemblies where the simple combinations of uranium fuel and diluent (graphite and stainless steel) in their core regions were systematically varied, the neutron spectra of them cover from the intermediate to fast one. Taking their advantages, benchmark models with respect to the central fission rate ratios had been recently developed for the evaluation of the TRU's fission cross sections. As an application of these benchmark models, the Japanese Evaluated Nuclear Data Library JENDL-4.0 was utilized by a Monte Carlo calculation code. Several results show large discrepancies between the calculation and experimental values. The benchmark models would be well suited for the evaluation and modification of the nuclear data for the TRU's fission cross sections.

  1. Nuclear fuel particles and method of making nuclear fuel compacts therefrom

    DOEpatents

    DeVelasco, Rubin I.; Adams, Charles C.

    1991-01-01

    Methods for making nuclear fuel compacts exhibiting low heavy metal contamination and fewer defective coatings following compact fabrication from a mixture of hardenable binder, such as petroleum pitch, and nuclear fuel particles having multiple layer fission-product-retentive coatings, with the dense outermost layer of the fission-product-retentive coating being surrounded by a protective overcoating, e.g., pyrocarbon having a density between about 1 and 1.3 g/cm.sup.3. Such particles can be pre-compacted in molds under relatively high pressures and then combined with a fluid binder which is ultimately carbonized to produce carbonaceous nuclear fuel compacts having relatively high fuel loadings.

  2. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    SciTech Connect

    Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

    2016-11-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  3. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    SciTech Connect

    Harp, Jason Michael; Stempien, John Dennis; Demkowicz, Paul Andrew

    2016-09-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  4. Feasibility of an antiproton-catalyzed fission-fragment rocket. Master's thesis

    SciTech Connect

    Hidinger, D.S.

    1992-03-01

    The purpose of this project was to investigate the feasibility of an antiproton catalyzed fission fragment rocket (FFR). The FFR is characterized by the extraction of fission fragments from the fuel, and the utilization of their kinetic energy for thrust generation. A significant drawback to previous FFR designs was the required critical nuclear pile as the fission fragment source. The author examined the possibility of replacing the critical pile with a sub-critical pile driven by antiprotons. Recent experiments have revealed that antiprotons stimulate highly energetic fissions in {sup 238}U, with a neutron multiplicity of 13.7 neutrons per fissions. This interaction was used as a throttled neutron source. The pile consisted of layers of fissile coated fibers which are designed to allow fission fragments to escape them, where the fragments collide with a fluid. The heated fluid is then ejected from the rocket to provide thrust. The calculations performed indicate that each antiproton injected into the pile can stimulate 8 or more fissions while maintaining a neutron multiplication of less than 0.4. Based on the results, the specific design presented was inadequate. Despite this, the concept of using the antiproton-U interaction as a source of thrust warrants further study.

  5. Analysis of Fission Products on the AGR-1 Capsule Components

    SciTech Connect

    Paul A. Demkowicz; Jason M. Harp; Philip L. Winston; Scott A. Ploger

    2013-03-01

    The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (i) the metal capsule components, (ii) the graphite fuel holders, (iii) the graphite spacers, and (iv) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs 134, Cs 137, Eu-154, and Sr 90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed to determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.2×10 2 (Capsule 3) to 3.8×10 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.

  6. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    SciTech Connect

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

  7. Fuel and cladding tests for fuel failure safety analysis. [LMFBR

    SciTech Connect

    Hunter, C.W.; Johnson, G.D.; Hu, W.L.; Cannon, N.S.; Feigenbutz, L.V.; Hinman, C.A.; Slagle, D.O.; Bard, F.E.

    1982-01-01

    Three systems for testing fuel pin cladding were developed to support the determination of the mode and location of cladding failure during transient overpower events or transient undercooled overpower (TUCOP) conditions. The TUCOP FCTT system consists of exposing cladding specimens to thermal and loading conditions typical of TUCOP events. The Mandrel Loading Test (MLT) system was designed to produce cladding deformation and failure by internal mechanical interaction loading of a heated cladding specimen. The Cladding Rip Propagation Test (CRPT) system measures the rip propagation behavior of cladding at different temperatures. Fuel deformation and fission gas release tests were performed to better understand fuel behavior. High temperature creep and hot pressing tests on mixed oxide fuel indicate that an enhancement in the creep rate occurs at temperatures above 2300/sup 0/C and that only a small proportion of the fabrication fuel porosity can be closed at temperatures above 2500/sup 0/C. Two fission gas induced modes of gross fuel behavior under transient thermal conditions have been demonstrated, i.e., brittle fracture along grain boundaries and massive plastic swelling of the fission-gas-containing fuel.

  8. Distribution of fission products palladium, silver, cerium and cesium in the un-corroded areas of the locally corroded SiC layer of a neutron irradiated TRISO fuel particle

    DOE PAGES

    Wen, Haiming; van Rooyen, Isabella J.

    2017-04-14

    Here, detailed electron microscopy studies were performed to investigate the distribution and composition of fission products in the SiC layer of a tristructural isotropic coated particle exhibiting localized corrosion. Previous studies on this particle indicated that pure carbon areas in the SiC layer, resulting from localized corrosion of SiC by Pd, provide pathways for Ag, Cd and Cs migration. This study reveals the presence of Ag- and/or Cd-containing precipitates in un-corroded SiC areas. Ag/Cd may exist by themselves or coexist with Pd. Ag/Cd mainly transport along SiC grain boundaries. An Ag-Pd-Cd precipitate was identified at a stacking fault inside amore » SiC grain, suggesting that intragranular transport of Ag/Cd is possible. Ce is present with Pd or Pd-U in some precipitates >50 nm. U and Ce frequently coexist with each other, whereas Ag/Cd usually does not coexist with U or Ce. No Cs was detected in any precipitates in the areas examined.« less

  9. The neutronics studies of fusion fission hybrid power reactor

    SciTech Connect

    Zheng Youqi; Wu Hongchun; Zu Tiejun; Yang Chao; Cao Liangzhi

    2012-06-19

    In this paper, a series of neutronics analysis of hybrid power reactor is proposed. The ideas of loading different fuels in a modular-type fission blanket is analyzed, fitting different level of fusion developments, i.e., the current experimental power output, the level can be obtained in the coming future and the high-power fusion reactor like ITER. The energy multiplication of fission blankets and tritium breeding ratio are evaluated as the criterion of design. The analysis is implemented based on the D-type simplified model, aiming to find a feasible 1000MWe hybrid power reactor for 5 years' lifetime. Three patterns are analyzed: 1) for the low fusion power, the reprocessed fuel is chosen. The fuel with high plutonium content is loaded to achieve large energy multiplication. 2) For the middle fusion power, the spent fuel from PWRs can be used to realize about 30 times energy multiplication. 3) For the high fusion power, the natural uranium can be directly used and about 10 times energy multiplication can be achieved.

  10. Status of fission yield data

    SciTech Connect

    England, T.R.; Blachot, J.

    1988-01-01

    In this paper we summarize the current status of the recent US evaluation for 34 fissioning nuclides at one or more neutron incident energies and for spontaneous fission. Currently there are 50 yields sets, and for each we have independent and cumulative yields and uncertainties for approximately 1100 fission products. When finalized the recommended data will become part of Version VI of the US ENDF/B. Other major evaluations in progress that are included in a recently formed IAEA Coordinated Research Program are also summarized. In a second part we review two empirical models in use to estimate independent yields. Comparison of model estimates with measured data is presented, including a comparison with some recent data obtained from Lohengrin (Cf-249 T). 18 refs., 13 figs., 3 tabs.

  11. Implementation of a new gamma spectrometer on the MERARG loop: Application to the volatile fission products release measurement

    SciTech Connect

    Bernard, S.; Gleizes, B.; Pontillon, Y.; Hanus, E.; Ducros, G.; Roure, C.

    2015-07-01

    The MERARG facility initially aims at the annealing of irradiated fuel samples to study the gaseous fission products release kinetics. In order to complete the evaluation of the source term potentially released during accidental situation, the MERARG experimental circuit has been enhanced with a new gamma spectrometer. This one is directly sighting the fuel and is devoted to the fission products release kinetics. Because of the specificities of the fuel measurements, it has been dimensioned and designed to match the specific requirements. The acquisition chain and the collimation system have been optimized for this purpose and a first set of two experiments have shown the good functioning of this new spectrometry facility. (authors)

  12. PRODUCING ENERGY AND RADIOACTIVE FISSION PRODUCTS

    DOEpatents

    Segre, E.; Kennedy, J.W.; Seaborg, G.T.

    1959-10-13

    This patent broadly discloses the production of plutonium by the neutron bombardment of uranium to produce neptunium which decays to plutonium, and the fissionability of plutonium by neutrons, both fast and thermal, to produce energy and fission products.

  13. Fission properties of the heaviest elements

    SciTech Connect

    Moller, P. |||; Nix, R.

    1995-03-01

    The authors discuss fission properties of the heaviest elements. In particular they focus on stability with respect to spontaneous fission and on the prospects of extending the region of known nuclei beyond the peninsula of currently known nuclides.

  14. Fission Matrix Capability for MCNP Monte Carlo

    NASA Astrophysics Data System (ADS)

    Brown, Forrest; Carney, Sean; Kiedrowski, Brian; Martin, William

    2014-06-01

    We describe recent experience and results from implementing a fission matrix capability into the MCNP Monte Carlo code. The fission matrix can be used to provide estimates of the fundamental mode fission distribution, the dominance ratio, the eigenvalue spectrum, and higher mode forward and adjoint eigenfunctions of the fission neutron source distribution. It can also be used to accelerate the convergence of the power method iterations and to provide basis functions for higher-order perturbation theory. The higher-mode fission sources can be used in MCNP to determine higher-mode forward fluxes and tallies, and work is underway to provide higher-mode adjoint-weighted fluxes and tallies. Past difficulties and limitations of the fission matrix approach are overcome with a new sparse representation of the matrix, permitting much larger and more accurate fission matrix representations. The new fission matrix capabilities provide a significant advance in the state-of-the-art for Monte Carlo criticality calculations.

  15. Effects of fission yield data in the calculation of antineutrino spectra for U235(n,fission) at thermal and fast neutron energies

    DOE PAGES

    Sonzogni, A. A.; McCutchan, E. A.; Johnson, T. D.; ...

    2016-04-01

    Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 235U fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of 86Ge generates an excess of antineutrinosmore » at 5–7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0–7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Lastly, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.« less

  16. Effects of Fission Yield Data in the Calculation of Antineutrino Spectra for 235U (n ,fission) at Thermal and Fast Neutron Energies

    NASA Astrophysics Data System (ADS)

    Sonzogni, A. A.; McCutchan, E. A.; Johnson, T. D.; Dimitriou, P.

    2016-04-01

    Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 235U 235 fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of 86Ge generates an excess of antineutrinos at 5-7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0-7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Finally, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.

  17. Effects of Fission Yield Data in the Calculation of Antineutrino Spectra for ^{235}U(n,fission) at Thermal and Fast Neutron Energies.

    PubMed

    Sonzogni, A A; McCutchan, E A; Johnson, T D; Dimitriou, P

    2016-04-01

    Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 ^{235}U fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of ^{86}Ge generates an excess of antineutrinos at 5-7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0-7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Finally, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.

  18. Fission Product Monitoring and Release Data for the Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John B. Walter; Jason M. Harp; Mark W. Drigert; Edward L. Reber

    2010-10-01

    The AGR-1 experiment is a fueled multiple-capsule irradiation experiment that was irradiated in the Advanced Test Reactor (ATR) from December 26, 2006 until November 6, 2009 in support of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Fuel Development and Qualification program. An important measure of the fuel performance is the quantification of the fission product releases over the duration of the experiment. To provide this data for the inert fission gasses(Kr and Xe), a fission product monitoring system (FPMS) was developed and implemented to monitor the individual capsule effluents for the radioactive species. The FPMS continuously measured the concentrations of various krypton and xenon isotopes in the sweep gas from each AGR-1 capsule to provide an indicator of fuel irradiation performance. Spectrometer systems quantified the concentrations of Kr-85m, Kr-87, Kr-88, Kr-89, Kr-90, Xe-131m, Xe-133, Xe 135, Xe 135m, Xe-137, Xe-138, and Xe-139 accumulated over repeated eight hour counting intervals.-. To determine initial fuel quality and fuel performance, release activity for each isotope of interest was derived from FPMS measurements and paired with a calculation of the corresponding isotopic production or birthrate. The release activities and birthrates were combined to determine Release-to-Birth ratios for the selected nuclides. R/B values provide indicators of initial fuel quality and fuel performance during irradiation. This paper presents a brief summary of the FPMS, the release to birth ratio data for the AGR-1 experiment and preliminary comparisons of AGR-1 experimental fuels data to fission gas release models.

  19. Magnesium transport extraction of transuranium elements from LWR fuel

    DOEpatents

    Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.; Pierce, R. Dean

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.

  20. Process for treating fission waste. [Patent application

    DOEpatents

    Rohrmann, C.A.; Wick, O.J.

    1981-11-17

    A method is described for the treatment of fission waste. A glass forming agent, a metal oxide, and a reducing agent are mixed with the fission waste and the mixture is heated. After melting, the mixture separates into a glass phase and a metal phase. The glass phase may be used to safely store the fission waste, while the metal phase contains noble metals recovered from the fission waste.

  1. Nuclear fission with diffusive dynamics

    NASA Astrophysics Data System (ADS)

    Cha, D.; Bertsch, G. F.

    1992-07-01

    We investigate the dynamics of nuclear fission, assuming purely diffusive motion up to the saddle point. The resulting Smoluchowski equation is solved for conditions appropriate to the 16O+142Nd-->158Er reaction at 207 MeV. The solution is characterized by an equilibration time τ0 for the system to reach steady state, and the fission decay rate in steady state, Λ. We find that the equilibration time τ0 plays a very small role in determining the number of prescission neutrons. The diffusion coefficient extracted from the experimental data is larger than the theoretical in the work of Bush, Bertsch, and Brown by a factor of 5-11.

  2. Search for Singlet Fission Chromophores

    SciTech Connect

    Havlas, Z.; Akdag, A.; Smith, M. B.; Dron, P.; Johnson, J. C.; Nozik, A. J.; Michl, J.

    2012-01-01

    Singlet fission, in which a singlet excited chromophore shares its energy with a ground-state neighbor and both end up in their triplet states, is of potential interest for solar cells. Only a handful of compounds, mostly alternant hydrocarbons, are known to perform efficiently. In view of the large number of conditions that a successful candidate for a practical cell has to meet, it appears desirable to extend the present list of high performers to additional classes of compounds. We have (i) identified design rules for new singlet fission chromophores and for their coupling to covalent dimers, (ii) synthesized them, and (iii) evaluated their performance as neat solids or covalent dimers.

  3. Ballistic piston fissioning plasma experiment.

    NASA Technical Reports Server (NTRS)

    Miller, B. E.; Schneider, R. T.; Thom, K.; Lalos, G. T.

    1971-01-01

    The production of fissioning uranium plasma samples such that the fission fragment stopping distance is less than the dimensions of the plasma is approached by using a ballistic piston device for the compression of uranium hexafluoride. The experimental apparatus is described. At room temperature the gun can be loaded up to 100 torr UF6 partial pressure, but at compression a thousand fold increase of pressure can be obtained at a particle density on the order of 10 to the 19th power per cu cm. Limited spectral studies of UF6 were performed while obtaining the pressure-volume data. The results obtained and their implications are discussed.

  4. Fission at intermediate neutron energies

    NASA Astrophysics Data System (ADS)

    Lo Meo, S.; Mancusi, D.; Massimi, C.; Vannini, G.; Ventura, A.

    2014-09-01

    In the present work, as a theoretical support to the campaign of neutron cross section measurements at the n_TOF facility at CERN[1], Monte Carlo calculations of fission induced by neutrons in the energy range from 100 MeV to 1 GeV are carried out by means of a recent version of the Liege Intranuclear Cascade Model, INCL++[6], coupled with different evaporation-fission codes, such as Gemini++[7] and ABLA07[8]. Theoretical cross sections are compared with experimental data obtained by the n_TOF collaboration and perspectives for future theoretical work are outlined.

  5. Analysis of a nuclear accident: fission and activation product releases from the Fukushima Daiichi nuclear facility as remote indicators of source identification, extent of release, and state of damaged spent nuclear fuel.

    PubMed

    Schwantes, Jon M; Orton, Christopher R; Clark, Richard A

    2012-08-21

    Researchers evaluated radionuclide measurements of environmental samples taken from the Fukushima Daiichi nuclear facility and reported on the Tokyo Electric Power Co. Website following the 2011 tsunami-initiated catastrophe. This effort identified Units 1 and 3 as the major source of radioactive contamination to the surface soil near the facility. Radionuclide trends identified in the soils suggested that: (1) chemical volatility driven by temperature and reduction potential within the vented reactors' primary containment vessels dictated the extent of release of radiation; (2) all coolant had likely evaporated by the time of venting; and (3) physical migration through the fuel matrix and across the cladding wall were minimally effective at containing volatile species, suggesting damage to fuel bundles was extensive. Plutonium isotopic ratios and their distance from the source indicated that the damaged reactors were the major contributor of plutonium to surface soil at the source, decreasing rapidly with distance from the facility. Two independent evaluations estimated the fraction of the total plutonium inventory released to the environment relative to cesium from venting Units 1 and 3 to be ∼0.002-0.004%. This study suggests significant volatile radionuclides within the spent fuel at the time of venting, but not as yet observed and reported within environmental samples, as potential analytes of concern for future environmental surveys around the site. The majority of the reactor inventories of isotopes of less volatile elements like Pu, Nb, and Sr were likely contained within the damaged reactors during venting.

  6. Method of fission heat flux determination from experimental data

    DOEpatents

    Paxton, Frank A.

    1999-01-01

    A method is provided for determining the fission heat flux of a prime specimen inserted into a specimen of a test reactor. A pair of thermocouple test specimens are positioned at the same level in the holder and a determination is made of various experimental data including the temperature of the thermocouple test specimens, the temperature of bulk water channels located in the test holder, the gamma scan count ratios for the thermocouple test specimens and the prime specimen, and the thicknesses of the outer clads, the fuel fillers, and the backclad of the thermocouple test specimen. Using this experimental data, the absolute value of the fission heat flux for the thermocouple test specimens and prime specimen can be calculated.

  7. Spontaneous fission properties of the heavy elements: Bimodal fission

    SciTech Connect

    Hulet, E.K.

    1988-11-11

    We have measured the mass and kinetic-energy distributions from the spontaneous fission of SVYFm, SVYNo, SVZMd, SWMd, SW(104), and SWSNo. All are observed to fission with a symmetrical division of mass, whereas the total-kinetic-energy (TKE) distributions strongly deviated from the Gaussian shape characteristically found in the fission of all other actinides. When the TKE distributions are resolved into two Gaussian's, the constituent peaks lie near 200 and near 233 MeV. We conclude two modes or bimodal fission is occurring in five of the six nuclides studied. Both modes are possible in the same nuclide, but one generally predominates. We also conclude the low-energy but mass-symmetrical mode is likely to extend to far heavier nuclei; while the high-energy mode will be restricted to a smaller region, a region of nuclei defined by the proximity of the fragments to the strong neutron and proton shells in TSSn. 21 refs., 7 figs., 1 tab.

  8. Cermet fuel reactors

    SciTech Connect

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  9. Stability Study of the RERTR Fuel Microstructure

    SciTech Connect

    Jian Gan; Dennis Keiser; Brandon Miller; Daniel Wachs

    2014-04-01

    The irradiation stability of the interaction phases at the interface of fuel and Al alloy matrix as well as the stability of the fission gas bubble superlattice is believed to be very important to the U-Mo fuel performance. In this paper the recent result from TEM characterization of Kr ion irradiated U-10Mo-5Zr alloy will be discussed. The focus will be on the phase stability of Mo2-Zr, a dominated second phase developed at the interface of U-10Mo and the Zr barrier in a monolithic fuel plate from fuel fabrication. The Kr ion irradiations were conducted at a temperature of 200 degrees C to an ion fluence of 2.0E+16 ions/cm2. To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated dispersion U-7Mo fuel and monolithic U-10Mo fuel, a FIB-TEM sample of the irradiated U-10Mo fuel (3.53E+21 fission/cm3) was used for a TEM in-situ heating experiment. The preliminary result showed extraordinary thermal stability of the fission gas bubble superlattice. The implication of the TEM observation from these two experiments on the fuel microstructural evolution under irradiation will be discussed.

  10. Pyrolytic carbon-coated nuclear fuel

    DOEpatents

    Lindemer, Terrence B.; Long, Jr., Ernest L.; Beatty, Ronald L.

    1978-01-01

    An improved nuclear fuel kernel having at least one pyrolytic carbon coating and a silicon carbon layer is provided in which extensive interaction of fission product lanthanides with the silicon carbon layer is avoided by providing sufficient UO.sub.2 to maintain the lanthanides as oxides during in-reactor use of said fuel.

  11. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1980-04-29

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

  12. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

  13. Fuel pin cladding

    DOEpatents

    Vaidyanathan, S.; Adamson, M.G.

    1983-12-16

    An improved fuel pin cladding, particularly adapted for use in breeder reactors, is described which consist of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel an/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

  14. Fuel pin cladding

    DOEpatents

    Vaidyanathan, S.; Adamson, M.G.

    1986-01-28

    Disclosed is an improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients. 2 figs.

  15. Fuel pin cladding

    DOEpatents

    Vaidyanathan, Swaminathan; Adamson, Martyn G.

    1986-01-01

    An improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

  16. Thermal Simulator Development: Non-Nuclear Testing of Space Fission Systems

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.; Dickens, Ricky E.

    2006-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power system. At the NASA MSFC Early Flight Fission Test Facility (EFF-TF), highly designed electric heaters are used to simulate the heat from nuclear fuel to test space fission power and propulsion systems. To allow early utilization, nuclear system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. In this test strategy, highly designed electric heaters are used to simulate the heat from nuclear fuel, allowing one to develop a significant understanding of individual components and integrated system operation without the cost, time and safety concerns associated with nuclear testing.

  17. Testing in Support of Fission Surface Power System Qualification

    NASA Technical Reports Server (NTRS)

    Houts, Mike; Bragg-Sitton, Shannon; Godfroy, Tom; Martin, Jim; Pearson, Boise; VanDyke, Melissa

    2007-01-01

    The strategy for qualifying a FSP system could have a significant programmatic impact. The US has not qualified a space fission power system since launch of the SNAP-10A in 1965. This paper explores cost-effective options for obtaining data that would be needed for flight qualification of a fission system. Qualification data could be obtained from both nuclear and non-nuclear testing. The ability to perform highly realistic nonnuclear testing has advanced significantly throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modern FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.) and extensive data to be taken from the core region. For transient testing, pin power during a transient is calculated based on the reactivity feedback that would occur given measured values of test article temperature and/or dimensional changes. The reactivity feedback coefficients needed for the test are either calculated or measured using cold/warm zero-power criticals. In this way non-nuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. FSP fuels and materials are typically chosen to ensure very high confidence in operation at design burnups, fluences, and temperatures. However, facilities exist (e.g. ATR, HFIR) for affordably performing in-pile fuel and materials irradiations, if such testing is desired. Ex-core materials and components (such as alternator materials, control drum drives, etc.) could be irradiated in university or DOE

  18. Molten fluoride fuel salt chemistry

    SciTech Connect

    Toth, L.M.; Del Cul, G.D.; Dai, S.; Metcalf, D.H.

    1994-09-01

    The chemistry of molten fluorides is traced from their development as fuels in the Molten Salt Reactor Experiment with important factors in their selection being discussed. Key chemical characteristics such as solubility, redox behavior, and chemical activity are explained as they relate to the behavior of molten fluoride fuel systems. Fission product behavior is described along with processing experience. Development requirements for fitting the current state of the chemistry to modern nuclear fuel system are described. It is concluded that while much is known about molten fluoride behavior, processing and recycle of the fuel components is a necessary factor if future systems are to be established.

  19. Gaseous fuel reactor research

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  20. Curved Waveguide Based Nuclear Fission for Small, Lightweight Reactors

    NASA Technical Reports Server (NTRS)

    Coker, Robert; Putnam, Gabriel

    2012-01-01

    advance could be applied to direct propulsion through guided fission products or as a secondary energy source for high impulse electric propulsion. It would help meet national needs for highly efficient energy sources with limited dependence on fossil fuels or conflict materials, and it would improve the use of low grade fissile materials which would help reduce national stockpiles and waste.

  1. Steady-state irradiation testing of U-Pu-Zr fuel to >18% burnup

    SciTech Connect

    Pahl, R.G.; Wisner, R.S. ); Billone, M.C.; Hofman, G.L. )

    1990-01-01

    Tests of austenitic stainless steel clad U-xP-10Zr fuel (x=o, 8, 19 wt. %) to peak burnups as high as 18.4 at. % have been completed in the EBR-II. Fuel swelling and fractional fission gas release are slowly increasing functions of burnup beyond 2 at. % burnup. Increasing plutonium content in the fuel reduces swelling and decreases the amount of fission gas which diffuses from fuel to plenum. LIFE-METAL code modelling of cladding strains is consistent with creep by fission gas loading and irradiation-induced swelling mechanisms. Fuel/cladding chemical interaction involves the ingress of rare-earth fission products. Constituent redistribution in the fuel had not limited steady-state performance. Cladding breach behavior at closure welds, in the gas plenum, and in the fuel column region have been benign events. 3 refs., 5 figs.

  2. Centromeric chromatin in fission yeast.

    PubMed

    Partridge, Janet F

    2008-05-01

    A fundamental requirement for life is the ability of cells to divide properly and to pass on to their daughters a full complement of genetic material. The centromere of the chromosome is essential for this process, as it provides the DNA sequences on which the kinetochore (the proteinaceous structure that links centromeric DNA to the spindle microtubules) assembles to allow segregation of the chromosomes during mitosis. It has long been recognized that kinetochore assembly is subject to epigenetic control, and deciphering how centromeres promote faithful chromosome segregation provides a fascinating intellectual challenge. This challenge is made more difficult by the scale and complexity of DNA sequences in metazoan centromeres, thus much research has focused on dissecting centromere function in the single celled eukaryotic yeasts. Interestingly, in spite of similarities in the genome size of budding and fission yeasts, they seem to have adopted some striking differences in their strategy for passing on their chromosomes. Budding yeast have "point" centromeres, where a 125 base sequence is sufficient for mitotic propagation, whereas fission yeast centromeres are more reminiscent of the large repetitive centromeres of metazoans. In addition, the centromeric heterochromatin which coats centromeric domains of fission yeast and metazoan centromeres and is critical for their function, is largely absent from budding yeast centromeres. This review focuses on the assembly and maintenance of centromeric chromatin in the fission yeast.

  3. Space Fission System Test Effectiveness

    SciTech Connect

    Houts, Mike; Schmidt, Glen L.; Van Dyke, Melissa; Godfroy, Tom; Martin, James; Bragg-Sitton, Shannon; Dickens, Ricky; Salvail, Pat; Harper, Roger

    2004-02-04

    Space fission technology has the potential to enable rapid access to any point in the solar system. If fission propulsion systems are to be developed to their full potential, however, near-term customers need to be identified and initial fission systems successfully developed, launched, and utilized. One key to successful utilization is to develop reactor designs that are highly testable. Testable reactor designs have a much higher probability of being successfully converted from paper concepts to working space hardware than do designs which are difficult or impossible to realistically test. ''Test Effectiveness'' is one measure of the ability to realistically test a space reactor system. The objective of this paper is to discuss test effectiveness as applied to the design, development, flight qualification, and acceptance testing of space fission systems. The ability to perform highly effective testing would be particularly important to the success of any near-term mission, such as NASA's Jupiter Icy Moons Orbiter, the first mission under study within NASA's Project Prometheus, the Nuclear Systems Program.

  4. Energetics of the fission process

    NASA Astrophysics Data System (ADS)

    Gönnenwein, Friedrich

    1994-09-01

    The mass asymmetry of fragments from nuclear fission of heavy nuclei is reviewed. While mass asymmetry is a common and well-known phenomenon for low-energy fission of the lighter actinides, more recent experiments have demonstrated that, for the heaviest actinides, the mass distribution switches to a symmetric one. On the other hand, it has been discovered that, though for fissioning nuclei with mass numbers A225 the mass distribution is basically symmetric, an asymmetric component is clearly to be identified for nuclei down to the Pb-region. In the absence of a generally accepted dynamical theory of fission, the above experimental findings are discussed in terms of static energy considerations. Triggered from the outset by the structure of the potential energy surface at the saddlepoint, the energy balance at the scission point between the available energy ( Q-value) of the reaction and the Coulomb and deformation energy of the nascent fragments is shown to steer the characteristics of the fragment mass distributions.

  5. Multimodal fission and neutron evaporation

    SciTech Connect

    Brosa, U.

    1988-10-01

    The average multiplicities nu-bar(A) of prompt neutrons emitted in the spontaneous fission of /sup 252/Cf and /sup 258/Fm are derived. Two new features are predicted: A simple sawtooth for /sup 258/Fm and a triple one for /sup 252/Cf. Experiments to check these predictions should be feasible now.

  6. Space Fission System Test Effectiveness

    NASA Astrophysics Data System (ADS)

    Houts, Mike; Schmidt, Glen L.; van Dyke, Melissa; Godfroy, Tom; Martin, James; Bragg-Sitton, Shannon; Dickens, Ricky; Salvail, Pat; Harper, Roger

    2004-02-01

    Space fission technology has the potential to enable rapid access to any point in the solar system. If fission propulsion systems are to be developed to their full potential, however, near-term customers need to be identified and initial fission systems successfully developed, launched, and utilized. One key to successful utilization is to develop reactor designs that are highly testable. Testable reactor designs have a much higher probability of being successfully converted from paper concepts to working space hardware than do designs which are difficult or impossible to realistically test. ``Test Effectiveness'' is one measure of the ability to realistically test a space reactor system. The objective of this paper is to discuss test effectiveness as applied to the design, development, flight qualification, and acceptance testing of space fission systems. The ability to perform highly effective testing would be particularly important to the success of any near-term mission, such as NASA's Jupiter Icy Moons Orbiter, the first mission under study within NASA's Project Prometheus, the Nuclear Systems Program.

  7. Etching fission tracks in zircons.

    PubMed

    Naeser, C W

    1969-07-25

    A new technique has been developed whereby fission tracks can be etched in zircon with a solution of sodiuim hydroxide at 220 degrees C. Etching time varied between 15 minutes and 5 houtrs. Colored zircon required less etching time than the colorless varieties.

  8. Neutronic fuel element fabrication

    DOEpatents

    Korton, George

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure

  9. A New Approach for the Permanent Disposal of Long Lived Fission Waste

    SciTech Connect

    Perkins, L J

    2007-03-27

    Nuclear fission can meet humanity's disparate requirements for carbon-free energy throughout this century and for millennia to come - not only for electricity but also as a source of hydrogen for transportation fuels and a heat source for desalination. However, most countries are not pursuing fission as an option for future energy and global climate needs. One paramount reason is diminished public acceptance over concerns of waste disposal. We would also add 'fuel resources' as a major future concern, because fission is not sustainable in the long term with the present 'once-through' fuel that utilizes less than 1% of the mined uranium and consigns its fertile potential to a permanent waste repository. Accordingly, global scale fission will become attainable (i.e., doable) if and when an integrated solution to this overall 'fuel-cycle' problem is realized. It is the back-end of the fuel cycle - i.e., the need for permanent storage of spent fuel and high-level waste - that has become the focus of much of the criticism. In particular, the construction and implementation of permanent waste repositories such as Yucca Mountain is becoming increasingly problematic from a financial and political perspective. The major shortcoming of these conventional repositories is that they must accommodate the whole spent fuel output from once-through fuel cycles. They are thus burdened with very large masses of material but where less than 1% is long-term, hazardous waste and where only a small fraction of the potential nuclear energy has been extracted. Second, such facilities must ensure integrity of waste containment for tens of thousands of years. Given that anything more than a few hundred years hence is unknowable and wholly unpredictable as far as future civilizations are concerned, public perception is that such facilities cannot be guaranteed to be absolutely secure for their envisaged lifetimes of tens of millennia.

  10. Development of a multiscale thermal conductivity model for fission gas in UO2

    NASA Astrophysics Data System (ADS)

    Tonks, Michael R.; Liu, Xiang-Yang; Andersson, David; Perez, Danielle; Chernatynskiy, Aleksandr; Pastore, Giovanni; Stanek, Christopher R.; Williamson, Richard

    2016-02-01

    Accurately predicting changes in the thermal conductivity of light water reactor UO2 fuel throughout its lifetime in reactor is an essential part of fuel performance modeling. However, typical thermal conductivity models from the literature are empirical. In this work, we begin to develop a mechanistic thermal conductivity model by focusing on the impact of gaseous fission products, which is coupled to swelling and fission gas release. The impact of additional defects and fission products will be added in future work. The model is developed using a combination of atomistic and mesoscale simulation, as well as analytical models. The impact of dispersed fission gas atoms is quantified using molecular dynamics simulations corrected to account for phonon-spin scattering. The impact of intragranular bubbles is accounted for using an analytical model that considers phonon scattering. The impact of grain boundary bubbles is determined using a simple model with five thermal resistors that are parameterized by comparing to 3D mesoscale heat conduction results. When used in the BISON fuel performance code to model four reactor experiments, it produces reasonable predictions without having been fit to fuel thermocouple data.

  11. Nuclear Fuel Reprocessing: U.S. Policy Development

    DTIC Science & Technology

    2006-11-29

    Nuclear Fuel Reprocessing: U.S. Policy Development Anthony Andrews Specialist in Industrial Engineering and Infrastructure Policy Resources, Science...separate and recover fissionable plutonium from irradiated nuclear fuel. In the early stage of commercial nuclear power, reprocessing was thought essential...to supplying nuclear fuel. Federally sponsored breeder reactor development included research into advanced reprocessing technology. Several

  12. Two neutron correlations in photo-fission

    NASA Astrophysics Data System (ADS)

    Dale, D. S.; Kosinov, O.; Forest, T.; Burggraf, J.; Stave, S.; Warren, G.; Starovoitova, V.

    2016-09-01

    A large body of experimental work has established the strong kinematical correlation between fission fragments and fission neutrons. Here, we report on the progress of investigations of the potential for strong two neutron correlations arising from the nearly back-to-back nature of the two fission fragments that emit these neutrons in the photo-fission process. In initial measurements, a pulsed electron linear accelerator was used to generate bremsstrahlung photons that impinged upon an actinide target, and the energy and opening angle distributions of coincident neutrons were measured using a large acceptance neutron detector array. A planned comprehensive set of measurements of two neutron correlations in the photo-fission of actinides is expected to shed light on several fundamental aspects of the fission process including the multiplicity distributions associated with the light and heavy fission fragments, the nuclear temperatures of the fission fragments, and the mass distribution of the fission fragments as a function of energy released. In addition to these measurements providing important nuclear data, the unique kinematics of fission and the resulting two neutron correlations have the potential to be the basis for a new tool to detect fissionable materials. A key technical challenge of this program arises from the need to perform coincidence measurements with a low duty factor, pulsed electron accelerator. This has motivated the construction of a large acceptance neutron detector array, and the development of data analysis techniques to directly measure uncorrelated two neutron backgrounds.

  13. Two neutron correlations in photo-fission

    SciTech Connect

    Dale, D. S.; Kosinov, O.; Forest, T.; Burggraf, J.; Stave, S.; Warren, G.; Starovoitova, V.

    2016-01-01

    A large body of experimental work has established the strong kinematical correlation between fission fragments and fission neutrons. Here, we report on the progress of investigations of the potential for strong two neutron correlations arising from the nearly back-to-back nature of the two fission fragments that emit these neutrons in the photo-fission process. In initial measurements, a pulsed electron linear accelerator was used to generate bremsstrahlung photons that impinged upon an actinide target, and the energy and opening angle distributions of coincident neutrons were measured using a large acceptance neutron detector array. A planned comprehensive set of measurements of two neutron correlations in the photo-fission of actinides is expected to shed light on several fundamental aspects of the fission process including the multiplicity distributions associated with the light and heavy fission fragments, the nuclear temperatures of the fission fragments, and the mass distribution of the fission fragments as a function of energy released. In addition to these measurements providing important nuclear data, the unique kinematics of fission and the resulting two neutron correlations have the potential to be the basis for a new tool to detect fissionable materials. A key technical challenge of this program arises from the need to perform coincidence measurements with a low duty factor, pulsed electron accelerator. This has motivated the construction of a large acceptance neutron detector array, and the development of data analysis techniques to directly measure uncorrelated two neutron backgrounds.

  14. Irradiation performance of full-length metallic IFR fuels

    SciTech Connect

    Tsai, H.; Neimark, L.A.

    1992-07-01

    An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II.

  15. Fission yield studies at the IGISOL facility

    NASA Astrophysics Data System (ADS)

    Penttilä, H.; Elomaa, V.-V.; Eronen, T.; Hakala, J.; Jokinen, A.; Kankainen, A.; Moore, I. D.; Rahaman, S.; Rinta-Antila, S.; Rissanen, J.; Rubchenya, V.; Saastamoinen, A.; Weber, C.; Äystö, J.

    2012-04-01

    Low-energy-particle-induced fission is a cost-effective way to produce neutron-rich nuclei for spectroscopic studies. Fission has been utilized at the IGISOL to produce isotopes for decay and nuclear structure studies, collinear laser spectroscopy and precision mass measurements. The ion guide technique is also very suitable for the fission yield measurements, which can be performed very efficiently by using the Penning trap for fission fragment identification and counting. The proton- and neutron-induced fission yield measurements at the IGISOL are reviewed, and the independent isotopic yields of Zn, Ga, Rb, Sr, Cd and In in 25MeV deuterium-induced fission are presented for the first time. Moving to a new location next to the high intensity MCC30/15 light-ion cyclotron will allow also the use of the neutron-induced fission to produce the neutron rich nuclei at the IGISOL in the future.

  16. Spent Nuclear Fuel Disposition

    SciTech Connect

    Wagner, John C.

    2016-05-22

    One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNF is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).

  17. Spent Nuclear Fuel Disposition

    DOE PAGES

    Wagner, John C.

    2016-05-22

    One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

  18. Nuclear Fuel Reprocessing

    SciTech Connect

    Harold F. McFarlane; Terry Todd

    2013-11-01

    Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore. Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor

  19. Fuel flexible fuel injector

    DOEpatents

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  20. Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

    NASA Astrophysics Data System (ADS)

    Barber, Duncan Henry

    During some postulated accidents at nuclear power stations, fuel cooling may be impaired. In such cases, the fuel heats up and the subsequent increased fission-gas release from the fuel to the gap may result in fuel sheath failure. After fuel sheath failure, the barrier between the coolant and the fuel pellets is lost or impaired, gases and vapours from the fuel-to-sheath gap and other open voids in the fuel pellets can be vented. Gases and steam from the coolant can enter the broken fuel sheath and interact with the fuel pellet surfaces and the fission-product inclusion on the fuel surface (including material at the surface of the fuel matrix). The chemistry of this interaction is an important mechanism to model in order to assess fission-product releases from fuel. Starting in 1995, the computer program SOURCE 2.0 was developed by the Canadian nuclear industry to model fission-product release from fuel during such accidents. SOURCE 2.0 has employed an early thermochemical model of irradiated uranium dioxide fuel developed at the Royal Military College of Canada. To overcome the limitations of computers of that time, the implementation of the RMC model employed lookup tables to pre-calculated equilibrium conditions. In the intervening years, the RMC model has been improved, the power of computers has increased significantly, and thermodynamic subroutine libraries have become available. This thesis is the result of extensive work based on these three factors. A prototype computer program (referred to as SC11) has been developed that uses a thermodynamic subroutine library to calculate thermodynamic equilibria using Gibbs energy minimization. The Gibbs energy minimization requires the system temperature (T) and pressure (P), and the inventory of chemical elements (n) in the system. In order to calculate the inventory of chemical elements in the fuel, the list of nuclides and nuclear isomers modelled in SC11 had to be expanded from the list used by SOURCE 2.0. A

  1. Prompt fission γ-ray data from spontaneous fission and the mechanism of fission-fragment de-excitation

    NASA Astrophysics Data System (ADS)

    Oberstedt, Stephan; Dragic, Aleksandar; Gatera, Angelique; Göök, Alf; Hambsch, Franz-Josef; Oberstedt, Andreas

    2017-09-01

    The investigation of prompt γ-ray emission in nuclear fission has a great relevance for the assessment of prompt heat generation in a reactor core and for the better understanding of the de-excitation mechanism of fission fragments. Some years ago experimental data was scarce and available only from a few fission reactions, 233,235U(nth, f), 239Pu(nth, f), and 252Cf(sf). Initiated by a high priority data request published by the OECD/NEA a dedicated prompt fission γ-ray measurement program is being conducted at the Joint Research Centre Geel. In recent years we obtained new and accurate prompt fission γ-ray spectrum (PFGS) characteristics (average number of photons per fission, average total energy per fission and mean photon energy) from 252Cf(sf), 235U(nth, f) and 239,241Pu(nth, f) within 2% of uncertainty. In order to understand the dependence of prompt fission γ-ray emission on the compound nuclear mass and excitation energy, we started a first measurement campaign on spontaneously fissioning plutonium and curium isotopes. Results on PFGS characteristics from 240,242Pu(sf) show a dependence on the fragment mass distribution rather than on the average prompt neutron multiplicity, pointing to a more complex competition between prompt fission γ-ray and neutron emission.

  2. Multidimensional Fuel Performance Code: BISON

    SciTech Connect

    2014-09-03

    BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phase field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.

  3. Importance functions in calculation of fuel burnup in a VVER

    SciTech Connect

    Semyonov, V. N.

    2011-12-15

    Equations that describe fuel burnup in a VVER are given. Equations for the neutron flux density and the content of fission products are presented in the canonical Cauchy form. Such form of representation of equations lends itself well for their use in solving problems related to optimization of the process of fuel burnup in a nuclear reactor. Also given are equations for the importance functions of neutrons and fission products that correspond to the basic system of equation for phase variables.

  4. BISON Theory Manual The Equations behind Nuclear Fuel Analysis

    SciTech Connect

    Hales, J. D.; Williamson, R. L.; Novascone, S. R.; Pastore, G.; Spencer, B. W.; Stafford, D. S.; Gamble, K. A.; Perez, D. M.; Liu, W.

    2016-09-01

    BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact, and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.

  5. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  6. A fission fragment detector for correlated fission output studies

    NASA Astrophysics Data System (ADS)

    Mosby, S.; Tovesson, F.; Couture, A.; Duke, D. L.; Kleinrath, V.; Meharchand, R.; Meierbachtol, K.; O'Donnell, J. M.; Perdue, B.; Richman, D.; Shields, D.

    2014-09-01

    A digital data acquisition system has been combined with a double Frisch gridded ionization chamber for use at both moderated and unmoderated neutron sources at the Los Alamos Neutron Science (LANSCE) facility. The high efficiency of the instrument combined with intense LANSCE beams and new acquisition system permits fission output measurements across 11 orders of magnitude incident neutron energy. The acquisition and analysis system is presented along with the first in-beam performance tests of the setup.

  7. DSP Algorithms for Fission Fragment and Prompt Fission Neutron Spectroscopy

    SciTech Connect

    Zeynalova, O.; Zeynalov, Sh.; Hambsch, F.-J.; Oberstedt, S.; Fabry, I.

    2009-10-29

    Digital signal processing (DSP) algorithms are in high demand for modern nuclear fission investigation due to importance of increase the accuracy of fissile nuclear data for new generation of nuclear power stations. DSP algorithms for fission fragment (FF) and prompt fission neutron (PFN) spectroscopy are described in the present work. The twin Frisch-grid ionization chamber (GTIC) is used to measure the kinetic energy-, mass- and angular distributions of the FF in the {sup 252}Cf(SF) reaction. Along with the neutron time-of-flight (TOF) measurement the correlation between neutron emission and FF mass and energy is investigated. The TOF is measured between common cathode of the GTIC and the neutron detector (ND) pulses. Waveform digitizers (WFD) having 12 bit amplitude resolution and 100 MHz sampling frequency are used for the detector pulse sampling. DSP algorithms are developed as recursive procedures to perform the signal processing, similar to those available in various nuclear electronics modules, such as constant fraction discriminator (CFD), pulse shape discriminator (PSD), peak-sensitive analogue-to-digital converter (pADC) and pulse shaping amplifier (PSA). To measure the angle between FF and the cathode plane normal to the GTIC a new algorithm is developed having advantage over the traditional analogue pulse processing schemes. Algorithms are tested by comparing the numerical simulation of the data analysis of the {sup 252}Cf(SF) reaction with data available from literature.

  8. Bright fission: singlet fission into a pair of emitting states.

    PubMed

    Casanova, David

    2015-06-09

    This paper reintroduces and explores the generation of two bright states from a single photon via a singlet fission mechanism in organic materials. This particular photophysical process is labeled here as bright fission (BF). The central part of the study is devoted to set the theoretical foundations of BF by discussing possible electronic mechanisms, the role of different excited states with various physical nature, the presence of competing deactivation channels, and the possible requirements for the BF viability. In a second part, some of the properties related to BF are computationally explored in anthracene. The analysis of computed high-lying excited states identifies several optical transitions as good candidates to trigger BF in anthracene. The approximation of excitonic couplings of these high energy levels to other electronic states within the same energy range suggests possible paths to populate electronic configurations potentially able to split in two independent spin singlets, i.e. singlet-singlet states. The study also explores the electronic structure of the energetically lowest singlet-singlet states in anthracene dimers and discusses the presence of charge transfer configurations and their relation to the singlet-singlet manifold. The computational results suggest fast relaxation to the lowest singlet-singlet state, from which the excitonic fission may occur. All in all, the present work aims at motivating to pursue further efforts in the study of the BF process in organic materials.

  9. Equilibrium Temperature Profiles within Fission Product Waste Forms

    SciTech Connect

    Kaminski, Michael D.

    2016-10-01

    We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.

  10. First Generation Least Expensive Approach to Fission (FiGLEAF) Testing Results

    NASA Technical Reports Server (NTRS)

    VanDyke, Melissa; Houts, Mike; Pedersen, Kevin; Godfroy, Tom; Dickens, Ricky; Poston, David; Reid, Bob; Salvail. Pat; Ring, Peter; Schmidt, George R. (Technical Monitor)

    2000-01-01

    Successful development of space fission systems will require an extensive program of affordable and realistic testing. In addition to tests related to design/development of the fission system, realistic testing of the actual flight unit must also be performed. Testing can be divided into two categories, non-nuclear tests and nuclear tests. Full power nuclear tests of space fission systems are expensive, time consuming, and of limited use, even in the best of programmatic environments. If the system is designed to operate within established radiation damage and fuel burn up limits while simultaneously being designed to allow close simulation of heat from fission using resistance heaters, high confidence in fission system performance and lifetime can be attained through a series of non-nuclear tests. Non-nuclear tests are affordable and timely, and the cause of component and system failures can be quickly and accurately identified. MSFC is leading a Safe Affordable Fission Engine (SAFE) test series whose ultimate goal is the demonstration of a 300 kW flight configuration system using non-nuclear testing. This test series is carried out in collaboration with other NASA centers, other government agencies, industry, and universities. The paper describes the SAFE test series, which includes test article descriptions, test results and conclusions, and future test plans.

  11. Comparison of predicted and measured fission product behavior in the Fort St. Vrain HTGR during the first three cycles of operation

    SciTech Connect

    Hanson, D.L.; Jovanovic, V.; Burnette, R.D.

    1985-10-01

    Fission product release from the reactor core has been predicted by the reference design methods and compared with reactor surveillance measurements and with the results of postirradiation examination (PIE) of spent FSV fuel elements. Overall, the predictive methods have been shown to be conservative: the predicted fission gas release at the end of Cycle 3 is about five times higher than observed. The dominant source of fission gas release is as-manufactured, heavy-metal contamination; in-service failure of the coated fuel particles appears to be negligible which is consistent with the PIE of spent fuel elements removed during the first two refuelings. The predicted releases of fission metals are insignificant compared to the release and subsequent decay of their gaseous precursors which is consistent with plateout probe measurements.

  12. The Fission Programme at the CERN n_TOF Facility

    NASA Astrophysics Data System (ADS)

    Tsinganis, A.; Barbagallo, M.; Berthoumieux, E.; Calviani, M.; Chiaveri, E.; Colonna, N.; Diakaki, M.; Duran, I.; Guerrero, C.; Gunsing, F.; Leal-Cidoncha, E.; Leong, L.-S.; Paradela, C.; Tarrio, D.; Tassan-Got, L.; Vlastou, R.

    Since 2001, the scientific programme of the CERN n_TOF facility has focused mainly on the study of radiative neutron capture reactions, which are of great interest to nuclear astrophysics and on neutron-induced fission reactions, which are of relevance for nuclear technology, as well as essential for the development of theoretical models of fission. In particular, taking advantage of the high instantaneous neutron flux and high energy resolution of the facility, as well as of high-performance detection and acquisition systems, accurate new measurements on several long-lived major and minor actinides, from 232Th to 245Cm, have been performed so far. Data on these isotopes are needed in order to improve the safety and efficiency of conventional reactors, as well as to develop new systems for nuclear energy production and treatment of nuclear waste, such as Generation IV reactors, Accelerator Driven Systems and reactors based on innovative fuel cycles. A review of the most important results on fission cross-sections and fragment properties obtained at n_TOF for a variety of (radioactive) isotopes is presented along with the perspectives arising from the coming on line in the second half of 2014 of a new 19 m flight-path, which will allow n_TOF to expand its measurement capabilities to even more rare or short-lived isotopes, such as 230Th, 232U, 238,240Pu and 244Cm.

  13. Research on stellarator-mirror fission-fusion hybrid

    NASA Astrophysics Data System (ADS)

    Moiseenko, V. E.; Kotenko, V. G.; Chernitskiy, S. V.; Nemov, V. V.; Ågren, O.; Noack, K.; Kalyuzhnyi, V. N.; Hagnestål, A.; Källne, J.; Voitsenya, V. S.; Garkusha, I. E.

    2014-09-01

    The development of a stellarator-mirror fission-fusion hybrid concept is reviewed. The hybrid comprises of a fusion neutron source and a powerful sub-critical fast fission reactor core. The aim is the transmutation of spent nuclear fuel and safe fission energy production. In its fusion part, neutrons are generated in deuterium-tritium (D-T) plasma, confined magnetically in a stellarator-type system with an embedded magnetic mirror. Based on kinetic calculations, the energy balance for such a system is analyzed. Neutron calculations have been performed with the MCNPX code, and the principal design of the reactor part is developed. Neutron outflux at different outer parts of the reactor is calculated. Numerical simulations have been performed on the structure of a magnetic field in a model of the stellarator-mirror device, and that is achieved by switching off one or two coils of toroidal field in the Uragan-2M torsatron. The calculations predict the existence of closed magnetic surfaces under certain conditions. The confinement of fast particles in such a magnetic trap is analyzed.

  14. Stochastic modelling of power reactor fuel behavior

    NASA Astrophysics Data System (ADS)

    Mirza, Shahid Nawaz

    An understanding of the in-reactor behavior of nuclear fuel is essential to the safe and economic operation of a nuclear power plant. It is no longer possible to achieve this without computer code calculations. A state of art computer code, FRODO, for Fuel ROD Operation, has been developed to model the steady state behavior of fuel pins in a light water reactor and to do sensitivity analysis. FRODO concentrates on the thermal performance, fission product release and pellet-clad interaction and can be used to predict the fuel failure under the prevailing conditions. FRODO incorporates the numerous uncertainties involved in fuel behavior modeling, using statistical methods, to ascertain fuel failures and their causes. Sensitivity of fuel failure to different fuel parameters and reactor conditions can be easily evaluated. FRODO has been used to analyze the sensitivities of fuel failures to coolant flow reductions. It is found that the uncertainties have pronounced effects on conclusions about fuel failures and their causes.

  15. The Pulsed Fission-Fusion (PUFF) Concept for Deep Space Exploration and Terrestrial Power Generation

    NASA Technical Reports Server (NTRS)

    Adams, Robert; Cassibry, Jason; Schillo, Kevin

    2017-01-01

    This team is exploring a modified Z-pinch geometry as a propulsion system, imploding a liner of liquid lithium onto a pellet containing both fission and fusion fuel. The plasma resulting from the fission and fusion burn expands against a magnetic nozzle, for propulsion, or a magnetic confinement system, for terrestrial power generation. There is considerable synergy in the concept; the lithium acts as a temporary virtual cathode, and adds reaction mass for propulsion. Further, the lithium acts as a radiation shield against generated neutrons and gamma rays. Finally, the density profile of the column can be tailored using the lithium sheath. Recent theoretical and experimental developments (e.g. tailored density profile in the fuel injection, shear stabilization, and magnetic shear stabilization) have had great success in mitigating instabilities that have plagued previous fusion efforts. This paper will review the work in evaluating the pellet sizes and z-pinch conditions for optimal PuFF propulsion. Trades of pellet size and composition with z-pinch power levels and conditions for the tamper and lithium implosion are evaluated. Current models, both theoretical and computational, show that a z-pinch can ignite a small (1 cm radius) fission-fusion target with significant yield. Comparison is made between pure fission and boosted fission targets. Performance is shown for crewed spacecraft for high speed Mars round trip missions and near interstellar robotic missions. The PuFF concept also offers a solution for terrestrial power production. PuFF can, with recycling of the effluent, achieve near 100% burnup of fission fuel, providing a very attractive power source with minimal waste. The small size of PuFF relative to today's plants enables a more distributed power network and less exposure to natural or man-made disruptions.

  16. Nuclear reactor fuel element with vanadium getter on cladding

    DOEpatents

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

  17. HTGR Fuel Technology Program. Semiannual report for the period ending March 31, 1983

    SciTech Connect

    Not Available

    1983-07-01

    This document reports the technical accomplishments of the HTGR Fuel Technology Program at GA Technologies Inc. during the first half of FY 83. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, core component verification, and core technology transfer tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR.

  18. Performance of breached LMFBR fuel pins during continued operation

    SciTech Connect

    Lambert, J.D.B.; Strain, R.V.; Gross, K.C.; Hofman, G.L.; Colburn, R.P.; Adamson, M.G.; Ukai, S.

    1985-01-01

    Four EBR-II tests were used to scope the behavior of breached mixed-oxide pins. After release of stored fission gas, delayed-neutron signals were large and easily detected, although not readily correlated with exposed fuel area. No problems were met during reactor operation or fuel handling. Fuel-sodium reaction caused only narrow breaches which released minute amounts of fuel and fission products; the reaction product appeared dense and non-friable. These initial results indicated LMFBR oxide pins could have considerable potential for operating in the breached mode.

  19. Modeling Deep Burn TRISO particle nuclear fuel

    NASA Astrophysics Data System (ADS)

    Besmann, T. M.; Stoller, R. E.; Samolyuk, G.; Schuck, P. C.; Golubov, S. I.; Rudin, S. P.; Wills, J. M.; Coe, J. D.; Wirth, B. D.; Kim, S.; Morgan, D. D.; Szlufarska, I.

    2012-11-01

    Under the DOE Deep Burn program TRISO fuel is being investigated as a fuel form for consuming plutonium and minor actinides, and for greater efficiency in uranium utilization. The result will thus be to drive TRISO particulate fuel to very high burn-ups. In the current effort the various phenomena in the TRISO particle are being modeled using a variety of techniques. The chemical behavior is being treated utilizing thermochemical analysis to identify phase formation/transformation and chemical activities in the particle, including kernel migration. Density functional theory is being used to understand fission product diffusion within the plutonia oxide kernel, the fission product's attack on the SiC coating layer, as well as fission product diffusion through an alternative coating layer, ZrC. Finally, a multiscale approach is being used to understand thermal transport, including the effect of radiation damage induced defects, in a model SiC material.

  20. Advanced waste forms from spent nuclear fuel

    SciTech Connect

    Ackerman, J.P.; McPheeters, C.C.

    1995-12-31

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed.

  1. Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 2: Power from nuclear fission

    NASA Technical Reports Server (NTRS)

    Clement, J. D.

    1973-01-01

    Different types of nuclear fission reactors and fissionable materials are compared. Special emphasis is placed upon the environmental impact of such reactors. Graphs and charts comparing reactor facilities in the U. S. are presented.

  2. Coherent singlet fission activated by symmetry breaking

    NASA Astrophysics Data System (ADS)

    Miyata, Kiyoshi; Kurashige, Yuki; Watanabe, Kazuya; Sugimoto, Toshiki; Takahashi, Shota; Tanaka, Shunsuke; Takeya, Jun; Yanai, Takeshi; Matsumoto, Yoshiyasu

    2017-10-01

    Singlet fission, in which a singlet exciton is converted to two triplet excitons, is a process that could be beneficial in photovoltaic applications. A full understanding of the dynamics of singlet fission in molecular systems requires detailed knowledge of the relevant potential energy surfaces and their (conical) intersections. However, obtaining such information is a nontrivial task, particularly for molecular aggregates. Here we investigate singlet fission in rubrene crystals using transient absorption spectroscopy and state-of-the-art quantum chemical calculations. We observe a coherent and ultrafast singlet-fission channel as well as the well-known and conventional thermally assisted incoherent channel. This coherent channel is accessible because the conical intersection for singlet fission on the excited-state potential energy surface is located very close to the equilibrium position of the ground-state potential energy surface and also because of the excitation of an intermolecular symmetry-breaking mode, which activates the electronic coupling necessary for singlet fission.

  3. MODELING AND FISSION CROSS SECTIONS FOR AMERICIUM.

    SciTech Connect

    ROCHMAN, D.; HERMAN, M.; OBLOZINSKY, P.

    2005-05-01

    This is the final report of the work performed under the LANL contract on the modeling and fission cross section for americium isotopes (May 2004-June 2005). The purpose of the contract was to provide fission cross sections for americium isotopes with the nuclear reaction model code EMPIRE 2.19. The following work was performed: (1) Fission calculations capability suitable for americium was implemented to the EMPIRE-2.19 code. (2) Calculations of neutron-induced fission cross sections for {sup 239}Am to {sup 244g}Am were performed with EMPIRE-2.19 for energies up to 20 MeV. For the neutron-induced reaction of {sup 240}Am, fission cross sections were predicted and uncertainties were assessed. (3) Set of fission barrier heights for each americium isotopes was chosen so that the new calculations fit the experimental data and follow the systematics found in the literature.

  4. Fission Product Transmutation in Mixed Radiation Fields

    SciTech Connect

    Harmon, Frank; Burgett, Erick; Starovoitova, Valeriia; Tsveretkov, Pavel

    2015-01-15

    Work under this grant addressed a part of the challenge facing the closure of the nuclear fuel cycle; reducing the radiotoxicity of lived fission products (LLFP). It was based on the possibility that partitioning of isotopes and accelerator-based transmutation on particular LLFP combined with geological disposal may lead to an acceptable societal solution to the problem of management. The feasibility of using photonuclear processes based on the excitation of the giant dipole resonance (GDR) by bremsstrahlung radiation as a cost effective transmutation method was accessed. The nuclear reactions of interest: (γ,xn), (n,γ), (γ,p) can be induced by bremsstrahlung radiation produced by high power electron accelerators. The driver of these processes would be an accelerator that produces a high energy and high power electron beam of ~ 100 MeV. The major advantages of such accelerators for this purpose are that they are essentially available “off the shelf” and potentially would be of reasonable cost for this application. Methods were examined that used photo produced neutrons or the bremsstrahlung photons only, or use both photons and neutrons in combination for irradiations of selected LLFP. Extrapolating the results to plausible engineering scale transmuters it was found that the energy cost for 129I and 99Tc transmutation by these methods are about 2 and 4%, respectively, of the energy produced from 1000MWe.

  5. MOX and MOX with 237Np/241Am Inert Fission Gas Generation Comparison in ATR

    SciTech Connect

    G. S. Chang; M. Robel; W. J. Carmack; D. J. Utterbeck

    2006-06-01

    The treatment of spent fuel produced in nuclear power generation is one of the most important issues to both the nuclear community and the general public. One of the viable options to long-term geological disposal of spent fuel is to extract plutonium, minor actinides (MA), and potentially long-lived fission products from the spent fuel and transmute them into short-lived or stable radionuclides in currently operating light-water reactors (LWR), thus reducing the radiological toxicity of the nuclear waste stream. One of the challenges is to demonstrate that the burnup-dependent characteristic differences between Reactor-Grade Mixed Oxide (RG-MOX) fuel and RG-MOX fuel with MA Np-237 and Am 241 are minimal, particularly, the inert gas generation rate, such that the commercial MOX fuel experience base is applicable. Under the Advanced Fuel Cycle Initiative (AFCI), developmental fuel specimens in experimental assembly LWR-2 are being tested in the northwest (NW) I-24 irradiation position of the Advanced Test Reactor (ATR). The experiment uses MOX fuel test hardware, and contains capsules with MOX fuel consisting of mixed oxide manufactured fuel using reactor grade plutonium (RG-Pu) and mixed oxide manufactured fuel using RG-Pu with added Np/Am. This study will compare the fuel neutronics depletion characteristics of Case-1 RG-MOX and Case-2 RG-MOX with Np/Am.

  6. RECOVERY OF ALUMINUM FROM FISSION PRODUCTS

    DOEpatents

    Blanco, R.E.; Higgins, I.R.

    1962-11-20

    A method is given for recovertng aluminum values from aqueous solutions containing said values together with fission products. A mixture of Fe/sub 2/O/ sub 3/ and MnO/sub 2/ is added to a solution containing aluminum and fission products. The resulting aluminum-containing supernatant is then separated from the fission product-bearing metal oxide precipitate and is contacted with a cation exchange resin. The aluminum sorbed on the resin is then eluted and recovered. (AEC)

  7. Increasing of MERARG experimental performances: on-line fission gas release measurement by mass spectrometry

    SciTech Connect

    Pontillon, Y.; Capdevila, H.; Clement, S.; Guigues, E.; Janulyte, A.; Zerega, Y.; Andre, J.

    2015-07-01

    The MERARG device - implemented at the LECASTAR Hot Laboratory, at the CEA Cadarache - allows characterizing nuclear fuels with respect to the behaviour of fission gases during thermal transients representative of normal or off normal operating nuclear power plant conditions. The fuel is heated in order to extract a part or the total gas inventory it contains. Fission Gas Release (FGR) is actually recorded by mean of both on-line gamma spectrometry station and micro gas chromatography. These two devices monitor the quantity and kinetics of fission gas release rate. They only address {sup 85}Kr radioactive isotope and the elemental quantification of Kr, Xe and He (with a relatively low detection limit in the latter case, typically 5-10 ppm). In order to better estimate the basic mechanisms that promote fission gas release from irradiated nuclear fuels, the CEA fuel study department decided to improve its experimental facility by modifying MERARG to extend the studies of gamma emitter fission gases to all gases (including Helium) with a complete isotopic distribution capability. To match these specifications, a Residual Gas Analyser (RGA) has been chosen as mass spectrometer. This paper presents a review of the main aspects of the qualification/calibration phase of the RGA type analyser. In particular, results recorded over three mass ranges 1-10 u, 80-90 u and 120-140 u in the two classical modes of MERARG, i.e. on-line and off-line measurements are discussed. Results obtained from a standard gas bottle show that the quantitative analysis at a few ppm levels can be achieved for all isotopes of Kr and Xe, as well as masses 2 and 4 u. (authors)

  8. FISSION PRODUCT REMOVAL FROM ORGANIC SOLUTIONS

    DOEpatents

    Moore, R.H.

    1960-05-10

    The decontamination of organic solvents from fission products and in particular the treatment of solvents that were used for the extraction of uranium and/or plutonium from aqueous acid solutions of neutron-irradiated uranium are treated. The process broadly comprises heating manganese carbonate in air to a temperature of between 300 and 500 deg C whereby manganese dioxide is formed; mixing the manganese dioxide with the fission product-containing organic solvent to be treated whereby the fission products are precipitated on the manganese dioxide; and separating the fission product-containing manganese dioxide from the solvent.

  9. Theoretical Description of the Fission Process

    SciTech Connect

    Witold Nazarewicz

    2003-07-01

    The main goals of the project can be summarized as follows: Development of effective energy functionals that are appropriate for the description of heavy nuclei. Our goal is to improve the existing energy density (Skyrme) functionals to develop a force that will be used in calculations of fission dynamics. Systematic self-consistent calculations of binding energies and fission barriers of actinide and trans-actinide nuclei using modern density functionals. This will be followed by calculations of spontaneous fission lifetimes and mass and charge divisions using dynamic adiabatic approaches based on the WKB approximation. Investigate novel microscopic (non-adiabatic) methods to study the fission process.

  10. METHOD FOR SEPARATING PLUTONIUM AND FISSION PRODUCTS EMPLOYING AN OXIDE AS A CARRIER FOR FISSION PRODUCTS

    DOEpatents

    Davies, T.H.

    1961-07-18

    Carrier precipitation processes for separating plutonium values from uranium fission products are described. Silicon dioxide or titanium dioxide in a finely divided state is added to an acidic aqueous solution containing hexavalent plutonium ions together with ions of uranium fission products. The supernatant solution containing plutonium ions is then separated from the oxide and the fission products associated therewith.

  11. A hemi-fission intermediate links two mechanistically distinct stages of membrane fission.

    PubMed

    Mattila, Juha-Pekka; Shnyrova, Anna V; Sundborger, Anna C; Hortelano, Eva Rodriguez; Fuhrmans, Marc; Neumann, Sylvia; Müller, Marcus; Hinshaw, Jenny E; Schmid, Sandra L; Frolov, Vadim A

    2015-08-06

    Fusion and fission drive all vesicular transport. Although topologically opposite, these reactions pass through the same hemi-fusion/fission intermediate, characterized by a 'stalk' in which only the outer membrane monolayers of the two compartments have merged to form a localized non-bilayer connection. Formation of the hemi-fission intermediate requires energy input from proteins catalysing membrane remodelling; however, the relationship between protein conformational rearrangements and hemi-fusion/fission remains obscure. Here we analysed how the GTPase cycle of human dynamin 1, the prototypical membrane fission catalyst, is directly coupled to membrane remodelling. We used intramolecular chemical crosslinking to stabilize dynamin in its GDP·AlF4(-)-bound transition state. In the absence of GTP this conformer produced stable hemi-fission, but failed to progress to complete fission, even in the presence of GTP. Further analysis revealed that the pleckstrin homology domain (PHD) locked in its membrane-inserted state facilitated hemi-fission. A second mode of dynamin activity, fuelled by GTP hydrolysis, couples dynamin disassembly with cooperative diminishing of the PHD wedging, thus destabilizing the hemi-fission intermediate to complete fission. Molecular simulations corroborate the bimodal character of dynamin action and indicate radial and axial forces as dominant, although not independent, drivers of hemi-fission and fission transformations, respectively. Mirrored in the fusion reaction, the force bimodality might constitute a general paradigm for leakage-free membrane remodelling.

  12. ORNL studies of fission product release under LWR accident conditions

    SciTech Connect

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

    1991-01-01

    High burnup Zircaloy-clad UO{sub 2} fuel specimens have been heated to study the release of fission products in tests simulating LWR accident conditions. The dominant variable was found to be temperature, with atmosphere, time, and burnup also being significant variables. Comparison of data from tests in steam and hydrogen, at temperatures of 2000 to 2700 K, have shown that the releases of the most volatile species (Kr, Xe, I, and Cs) are relatively insensitive to atmosphere. The releases of the less-volatile species (Sr, Mo, Ru, Sb, Te, Ba, and Eu), however, may vary by orders of magnitude depending on atmosphere. In addition, the atmosphere may drastically affect the mode and extent of fuel destruction.

  13. COMPARTMENTED REACTOR FUEL ELEMENT

    DOEpatents

    Cain, F.M. Jr.

    1962-09-11

    A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

  14. The chemical constitution of the fuel-clad gap in oxide fuel pins for nuclear reactors

    NASA Astrophysics Data System (ADS)

    Ball, R. G. J.; Burns, W. G.; Henshaw, J.; Mignanelli, M. A.; Potter, P. E.

    1989-09-01

    An assessment of the chemical equilibria which can occur in the fuel-clad gap of irradiated pins for water-cooled thermal reactors and liquid-sodium-cooled fast breeder reactors is presented. An understanding of such reactions is essential to the elucidation of mechanisms of cladding failure due to fission product elements. Particular attention is given to the variation of the chemical constitution of phases containing the fission product elements caesium, iodine and tellurium. The perturbation by irradiation of the thermal equilibrium concentrations of the gaseous species of these fission product elements has also been assessed in this paper.

  15. Feasibility study of a fission-suppressed tokamak fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Neef, W.S.; Berwald, D.H.; Garner, J.K.; Whitley, R.H.; Ghoniem, N.; Wong, C.P.C.; Maya, I.; Schultz, K.R.

    1984-12-01

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m/sup 2/ and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of /sup 233/U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U/sub 3/O/sub 8/ depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management.

  16. Design of pellet surface grooves for fission gas plenum

    SciTech Connect

    Carter, T.J.; Jones, L.R.; Macici, N.; Miller, G.C.

    1986-01-01

    In the Canada deuterium uranium pressurized heavy water reactor, short (50-cm) Zircaloy-4 clad bundles are fueled on-power. Although internal void volume within the fuel rods is adequate for the present once-through natural uranium cycle, the authors have investigated methods for increasing the internal gas storage volume needed in high-power, high-burnup, experimental ceramic fuels. This present work sought to prove the methodology for design of gas storage volume within the fuel pellets - specifically the use of grooves pressed or machined into the relatively cool pellet/cladding interface. Preanalysis and design of pellet groove shape and volume was accomplished using the TRUMP heat transfer code. Postirradiation examination (PIE) was used to check the initial design and heat transfer assumptions. Fission gas release was found to be higher for the grooved pellet rods than for the comparison rods with hollow or unmodified pellets. This had been expected from the initial TRUMP thermal analyses. The ELESIM fuel modeling code was used to check in-reactor performance, but some modifications were necessary to accommodate the loss of heat transfer surface to the grooves. It was concluded that for plenum design purposes, circumferential pellet grooves could be adequately modeled by the codes TRUMP and ELESIM.

  17. Data summary report for fission product release Test VI-7

    SciTech Connect

    Osborne, M.F.; Lorentz, R.A.; Travis, J.R.; Collins, J.L.; Webster, C.S.

    1995-05-01

    Test VI-7 was the final test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the Monticello boiling water reactor (BWR). The fuel had experienced a burnup of {approximately}-40 Mwd/kg U. It was heated in an induction furnace for successive 20-min periods at 2000 and 2300 K in a moist air-helium atmosphere. Integral releases were 69% for {sup 85}Kr, 52% for {sup 125}Sb, 71% for both {sup 134}Cs and {sup 137}Cs, and 0.04% for {sup 154}Eu. For the non-gamma-emitting species, release values for 42% for I, 4.1% for Ba, 5.3% for Mo, and 1.2% for Sr were determined. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.89 g, with 37% being collected on the thermal gradient tubes and 63% downstream on filters. Posttest examination of the fuel specimen indicated that most of the cladding was completely oxidized to ZrO{sub 2}, but that oxidation was not quite complete at the upper end. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL-Booth Model.

  18. VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

    DOEpatents

    Furgerson, W.T.

    1963-12-17

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

  19. Nuclear fuel elements and method of making same

    DOEpatents

    Schweitzer, Donald G.

    1992-01-01

    A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

  20. Assessment of improved fission-product transport models in VICTORIA against the ORNL HI and VI tests

    SciTech Connect

    Domagala, P.; Rest, J.; Zawadzki, S.A.

    1991-01-01

    New models for the release of fission-products from fuel have recently been incorporated into a comprehensive code, VICTORIA, for the prediction of radionuclide behavior under severe reactor accident conditions as an alternative to a simple Booth diffusion calculation. It is widely known that the Booth model has severe limitations when used under conditions of changing temperature and power. A new transport model based on a two node diffusive flow formulation has been implemented into VICTORIA. In addition to the diffusive flow model, other mechanisms such as grain growth, grain boundary sweeping and intergranular bubble behavior are taken into account. These physically based models focus on fission-product behavior in intact fuel geometries. While the VICTORIA program is primarily concerned with the behavior of severely degraded fuel geometries, the capability for accurately characterizing fission-product release from intact fuel must be demonstrated before other geometries can be reliably modeled. Results of VICTORIA simulations are compared with the results of in-cell heating tests on irradiated fuel. The fission-product transport models involved in these simulations are assessed and their predictive capability examined. 10 refs., 7 figs., 2 tabs.