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Sample records for fast reactor burnup

  1. Burnup concept for a long-life fast reactor core using MCNPX.

    SciTech Connect

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  2. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    SciTech Connect

    Widiawati, Nina Su’ud, Zaki

    2015-09-30

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  3. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    NASA Astrophysics Data System (ADS)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  4. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    SciTech Connect

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Asiah, Nur; Shafii, M. Ali; Khairurrijal

    2010-12-23

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (k{sub eff}) is in almost linear relations with the change of the fuel volume to coolant ratio.

  5. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

  6. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    SciTech Connect

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-22

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  7. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    NASA Astrophysics Data System (ADS)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

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

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

    SciTech Connect

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

    2015-09-30

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

  10. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    SciTech Connect

    Su'ud, Zaki; Sekimoto, H.

    2014-09-30

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  11. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  12. A Modal Expansion Equilibrium Cycle Perturbation Method for Optimizing High Burnup Fast Reactors

    NASA Astrophysics Data System (ADS)

    Touran, Nicholas W.

    This dissertation develops a simulation tool capable of optimizing advanced nuclear reactors considering the multiobjective nature of their design. An Enhanced Equilibrium Cycle (EEC) method based on the classic equilibrium method is developed to evaluate the response of the equilibrium cycle to changes in the core design. Advances are made in the consideration of burnup-dependent cross sections and dynamic fuel performance (fission gas release, fuel growth, and bond squeeze-out) to allow accuracy in high-burnup reactors such as the Traveling Wave Reactor. EEC is accelerated for design changes near a reference state through a new modal expansion perturbation method that expands arbitrary flux perturbations on a basis of λ-eigenmodes. A code is developed to solve the 3-D, multigroup diffusion equation with an Arnoldi-based solver that determines hundreds of the reference flux harmonics and later uses these harmonics to determine expansion coefficients required to approximate the perturbed flux. The harmonics are only required for the reference state, and many substantial and localized perturbations from this state are shown to be well-approximated with efficient expressions after the reference calculation is performed. The modal expansion method is coupled to EEC to produce the later-in-time response of each design perturbation. Because the code determines the perturbed flux explicitly, a wide variety of core performance metrics may be monitored by working within a recently-developed data management system called the ARMI. Through ARMI, the response of each design perturbation may be evaluated not only for the flux and reactivity, but also for reactivity coefficients, thermal hydraulics parameters, economics, and transient performance. Considering the parameters available, an automated optimization framework is designed and implemented. A non-parametric surrogate model using the Alternating Conditional Expectation (ACE) algorithm is trained with many design

  13. Microstructural Characterization of High Burn-up Mixed Oxide Fast Reactor Fuel

    SciTech Connect

    Melissa C. Teague; Brian P. Gorman; Steven L. Hayes; Douglas L. Porter; Jeffrey King

    2013-10-01

    High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.

  14. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    NASA Astrophysics Data System (ADS)

    Uwaba, Tomoyuki; Ito, Masahiro; Maeda, Koji

    2011-09-01

    The C3M irradiation test, which was conducted in the experimental fast reactor, "Joyo", demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, "Monju". The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 137Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  15. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    NASA Astrophysics Data System (ADS)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1- y Pu y )O2- x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance

  16. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    NASA Astrophysics Data System (ADS)

    Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

    2013-10-01

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  17. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

    SciTech Connect

    Uwaba, Tomoyuki; Ito, Masahiro; Mizuno, Tomoyasu; Katsuyama, Kozo; Makenas, Bruce J.; Wootan, David W.; Carmack, Jon

    2011-06-16

    The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39E26 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

  18. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

    SciTech Connect

    Tomoyuki Uwaba; Masahiro Ito; Kozo Katsuyama; Bruce J. Makenas; David W. Wootan; Jon Carmack

    2011-05-01

    The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 1026 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

  19. Inherent safety of minimum-burnup breed and burn reactors

    SciTech Connect

    Qvist, S.; Reenspan, E.

    2012-07-01

    Reactors that aim to sustain the breed and burn (B and B) mode of operation at minimum discharge burnup require excellent neutron economy, Minimum-burnup B and B cores are generally large and feature low neutron leakage probability and a hard neutron spectrum. While highly promising fuel cycles can be achieved with such designs, the very same features are pushing the limits of the core's ability to passively respond safely to unprotected accidents. Low leakage minimum-burnup sodium-cooled B and B cores have a large positive coolant void-worth and coolant temperature reactivity coefficient. In this study, the applicability of major approaches for fast reactor void-worth reduction is evaluated specifically for B and B cores. The design, shuffling scheme and performance of a new metallic-fueled, sodium-cooled minimum burnup B and B core, used as basis for the void-worth reduction analysis, is presented. The analysis shows that reactivity control systems based on passive {sup 6}Li injection during temperature excursions are the only option able to provide negative void-worth without significantly increasing the minimum burnup required for sustaining the B and B mode of operation. A new type of lithium expansion module (LEM) system was developed specifically for B and B cores and its effect on core performance is presented. (authors)

  20. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    SciTech Connect

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-09-30

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.

  1. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    NASA Astrophysics Data System (ADS)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  2. PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

    DOEpatents

    Puechl, K.H.

    1963-09-24

    A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

  3. Effects of Burnup and Temperature Distributions to CANDLE Burnup of Block-Type High Temperature Gas Cooled Reactor

    SciTech Connect

    Yasunori Ohoka; Ismile; Hiroshi Sekimoto

    2004-07-01

    The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top or from top to bottom of the core and without any change in their shapes. It can be applied easily to the block-type high temperature gas cooled reactor using an appropriate burnable poison mixed with uranium oxide fuel. In the present study, the burnup distribution and the temperature distribution in the core are investigated and their effects on the CANDLE burnup core characteristics are studied. In this study, the natural gadolinium is used as the burnable poison. With the fuel enrichment of 15%, the natural gadolinium concentration of 3.0% and the fuel pin pitch of 6.6 cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half width of power density distribution of 1.5 m for uniform group constant case at 900 K. When the effect of nuclide change by burnup is considered, the burning region speed becomes 25 cm/year and the axial half-width of power density distribution becomes 1.25 m. When the temperature distributions effect is considered, the effects on the core characteristics are smaller than the burnup distribution effect. The maximum fuel temperature of the parallel flow case is higher than the counter flow case. (authors)

  4. Fast Breeder Reactor studies

    SciTech Connect

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  5. The Integral Fast Reactor

    SciTech Connect

    Till, C.E.; Chang, Y.I. ); Lineberry, M.J. )

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

  6. The Integral Fast Reactor

    SciTech Connect

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab.

  7. Burn-up and neutron economy of accelerator-driven reactor

    SciTech Connect

    Takahashi, H.; Yang, W.; An, Y.; Yamazaki, Y.

    1997-07-01

    It is desirable to have only a small reactivity change in the large burn-up of a solid fuel fast reactor, so that the number of replacements or shuffling of the fuel can be reduced, and plant factor accordingly increased. Also, this reduces the number of control rods needed for the change in burn-up reactivity. In subcritical operation, power controlled by beam power is suggested, but this practice is not as economical as the use of control rods and makes more careful operation of the accelerator is required due to changes in the wake field. In subcritical operation, even a slightly subcritical one, the safety problems associated with a hard neutron spectrum can be alleviated. Neutron leakage from a flattened core, which is needed for operation of the critical fast reactor can be lessen by using the non flat core which has good neutron economy. For generating nuclear energy, it is essential to have a high neutron economy, although breeding the fuel is not welcomed in the present political climate, as is needed for transmuting long lived fission products. In contrast to the breeder, the accelerator driven reactor can separate the energy production from fuel production and processing. Thus, it is suited for non-proliferation of nuclear material by prohibiting the processing and production of fuel in the unrestricted area so this can be only done in international controlled areas which are restricted and remote.

  8. Integral Fast Reactor concept

    SciTech Connect

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

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  9. Fast quench reactor method

    SciTech Connect

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.; Berry, Ray A.

    1999-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

  10. Fast quench reactor method

    DOEpatents

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

    1999-08-10

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

  11. Fast Reactor Technology Preservation

    SciTech Connect

    Wootan, David W.; Omberg, Ronald P.

    2008-01-11

    There is renewed worldwide interest in developing and implementing a new generation of advanced fast reactors. International cooperative efforts are underway such as the Global Nuclear Energy Partnership (GNEP). Advanced computer modeling and simulation efforts are a key part of these programs. A recognized and validated set of Benchmark Cases are an essential component of such modeling efforts. Testing documentation developed during the operation of the Fast Flux Test Facility (FFTF) provide the information necessary to develop a very useful set of Benchmark Cases.

  12. CANDLE: The New Burnup Strategy

    SciTech Connect

    Sekimoto, Hiroshi; Ryu, Kouichi; Yoshimura, Yoshikane

    2001-11-15

    The new burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) is proposed. With this burnup strategy, distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes. The excess reactivity is constant during the burnup. Therefore, any control mechanisms for the burnup are not required. Calculation procedures are presented to find these shapes and the speed of the burning region with the neutron multiplication factor of a reactor employing this burnup strategy.To demonstrate the CANDLE burnup strategy, it is applied to a fast reactor with excellent neutron economy. Only the initially built reactor requires some fissile material such as plutonium or enriched uranium for the nuclear ignition region of its core, but only natural uranium or depleted uranium is required for the other region. Succeeding reactors require only natural or depleted uranium since the burning region of the previous reactor can be utilized for the ignition region. The life of a reactor can be made longer by elongating the core height. The drift speed of the burning region for the presented fast reactor design is {approx}4 cm/yr, which is a preferable value for designing a long-life reactor. The burnup of spent fuel is {approx}40%. It is equivalent to 40% utilization of natural uranium without reprocessing and enrichment.

  13. History of fast reactor fuel development

    NASA Astrophysics Data System (ADS)

    Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

    1993-09-01

    The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

  14. Spatial kinetics in fast reactors

    NASA Astrophysics Data System (ADS)

    Seleznev, E. F.; Belov, A. A.; Panova, I. S.; Matvienko, I. P.; Zhukov, A. M.

    2013-12-01

    The analysis of the solution to the spatial nonstationary equation of neutron transport is presented by the example of a fast reactor. Experiments in spatial kinetics conducted recently at the complex of critical assemblies (fast physical stand) and computations of their data using the TIMER code (for solving the nonstationary equation in multidimensional diffusion approximation for direct and inverse problems of reactor kinetics) have shown that kinetics of fast reactors substantially differs from kinetics of thermal reactors. The difference is connected with influence of the delayed neutron spectrum on rates of the process in a fast reactor.

  15. FAST NEUTRONIC REACTOR

    DOEpatents

    Snell, A.H.

    1957-12-01

    This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

  16. Fuel systems for compact fast space reactors

    SciTech Connect

    Cox, C.M.; Dutt, D.S.; Karnesky, R.A.

    1983-12-01

    About 200 refractory metal clad ceramic fuel pins have been irradiated in thermal reactors under the 1200 K to 1550 K cladding temperature conditions of primary relevance to space reactors. This paper reviews performance with respect to fissile atom density, operating temperatures, fuel swelling, fission gas release, fuel-cladding compatibility, and consequences of failure. It was concluded that UO/sub 2/ and UN fuels show approximately equal performance potential and that UC fuel has lesser potential. W/Re alloys have performed quite well as cladding materials, and Ta, Nb, and Mo/Re alloys, in conjunction with W diffusion barriers, show good promise. Significant issues to be addressed in the future include high burnup swelling of UN, effects of UO/sub 2/-Li coolant reaction in the event of fuel pin failure, and development of an irradiation performance data base with prototypically configured fuel pins irradiated in a fast neutron flux.

  17. 60Co as AN On-Line Burnup Indicator for Multi-Pass Pebble Bed Reactors

    NASA Astrophysics Data System (ADS)

    Hawari, Ayman I.; Chen, Jianwei

    2003-06-01

    Multi-pass pebble bed reactor concepts are characterized by circulating fuel systems that cycle the pebbles in and out of the core until the burnup limit is reached. Currently modular designs of such reactors, with nominal powers of approximately 300 MW-thermal, are under consideration for deployment internationally. A concern of the proposed designs is the ability to perform online measurements of the fuel burnup to determine whether a pebble has reached its end-of-life burnup limit (~ 80,000 MWD/MTU). In this work, computational simulations were performed to assess the utilization of a passive gamma ray spectrometric approach to perform this task. However, in addition to using the inherent signatures of the irradiated fuel, the use of the 59Co(n,γ)60Co reaction as a burnup indicator is considered. The results show that the activity ratio of 134Cs/60Co can provide an indicator that is accurate to within 5% at burnup greater than 20,000 MWD/MTU as the power is varied between 50% and 200% of the reactor's thermal power.

  18. Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

    NASA Astrophysics Data System (ADS)

    Sloma, Tanya Noel

    When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light

  19. Investigation on using neutron counting techniques for online burnup monitoring of pebble bed reactor fuels

    NASA Astrophysics Data System (ADS)

    Zhao, Zhongxiang

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor. This project investigated the feasibility of using the passive neutron counting and active neutron/gamma counting for the on line fuel burnup measurement for MPBR. To investigate whether there is a correlation between neutron emission and fuel burnup, the MPBR fuel depletion was simulated under different irradiation conditions by ORIGEN2. It was found that the neutron emission from an irradiated pebble increases with burnup super-linearly and reaches to 104 neutron/sec/pebble at the discharge burnup. The photon emission from an irradiated pebble was found to be in the order of 1013 photon/sec/pebble at all burnup levels. Analysis shows that the neutron emission rate of an irradiated pebble is sensitive to its burnup history and the spectral-averaged one-group cross sections used in the depletion calculations, which consequently leads to large uncertainty in the correlation between neutron emission and burnup. At low burnup levels, the uncertainty in the neutron emission/burnup correlation is too high and the neutron emission rate is too low so that it is impossible to determine a pebble's burnup by on-line neutron counting at low burnup levels. At high burnup levels, the uncertainty in the neutron emission rate becomes less but is still large in quantity. However, considering the super-linear feature of the correlation, the uncertainty in burnup determination was found to be ˜7% at the discharge burnup, which is acceptable. Therefore, total neutron emission rate of a pebble can be used as a burnup indicator to determine whether a pebble should be discharged or not. The feasibility of using passive neutron counting methods for the on-line burnup measurement was investigated by using a general Monte Carlo code, MCNP, to assess the detectability of the neutron emission and the capability to discriminate gamma noise by commonly used neutron detectors. It was found that both He-3

  20. Implications of Fast Reactor Transuranic Conversion Ratio

    SciTech Connect

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays

    2010-11-01

    Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 (“burners”) do not have blankets; the cases above CR=1 (“breeders”) have breeding blankets. The burnup was allowed to float while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is “attractive” for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR<1, the heat, gamma, and neutron emission increase as material is recycled. The uranium utilization is at or below 1%, just as it is in thermal reactors as both types of reactors require continuing fissile support. For CR>1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially found

  1. Fast reactors and nuclear nonproliferation

    SciTech Connect

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  2. Fast quench reactor and method

    DOEpatents

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

    1998-05-12

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

  3. The burnup dependence of light water reactor spent fuel oxidation

    SciTech Connect

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies

  4. Extended Burnup Demonstration Reactor Fuels Program. Annual progress report, April 1983-March 1984. [BWR

    SciTech Connect

    Exarhos, C.A.

    1985-06-20

    The US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities Nuclear Corporation have participated since 1979 in a cooperative Extended Burnup Demonstration Program. Under the program, standard ENC-fabricated reload fuel in the Big Rock Point and Oyster Creek reactor cores has been irradiated to discharge burnups at or beyond 35,000 MWD/MTU, one to two cycles beyond its originally projected exposure life. The program provides for examination of the fuel at poolside before and after each extended burnup cycle as well as for limited destructive hot cell examination. The 1984 progress report covers work performed under the EBD program between April 1983 and March 1984. Major milestones reached during the period include completion of a hot cell examination on four high burnup rods from Big Rock Point and of a poolside on the Oyster Creek EBD fuel at discharge. The hot cell examination of four rods at burnups to 37.2 GWD/MTU confirmed poolside measurements on the same fuel, showing the urania and gadolinia-bearing fuel rods to be in excellent condition. No major cladding degradation, pellet restructuring, or pellet-clad interaction was found in any of the samples examined. The Oyster Creek fuel, examined at an assembly average exposure of 34.5 GWD/MTU, showed good performance with regard to both diametral creepdown and clad oxide accumulation.

  5. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect

    Venkataraman, M.; Natarajan, R.; Raj, Baldev

    2007-07-01

    The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR) spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)

  6. Preliminary Study of Burnup Characteristics for a Simplified Small Pebble Bed Reactor

    NASA Astrophysics Data System (ADS)

    Irwanto, Dwi; Kato, Yukikata; Yamanaka, Ichiro; Obara, Toru

    2010-06-01

    Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu à Peu (little by little) fueling scheme. In the Peu à Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. The objectives of the the present study were to find a means of carrying out the exact calculations needed to analyze the Peu à Peu fuel-loading scheme and to optimize the fuel composition, and fuel-loading scheme to achieve better burnup characteristics. The Monte Carlo method is used to perform calculations with high accuracy. Before the calculation of the whole core, the analysis for the infinite geometry was performed. The power generated per mass consumed for each combination of the uranium enrichment and packing fraction was analyzed from the parametric survey. By using the optimal value obtained, a whole-core calculation for the small 20 MWth reactor was performed and the criticality and burnup of this design was analyzed.

  7. Preliminary Study of Burnup Characteristics for a Simplified Small Pebble Bed Reactor

    SciTech Connect

    Irwanto, Dwi; Kato, Yukikata; Obara, Toru; Yamanaka, Ichiro

    2010-06-22

    Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu a Peu (little by little) fueling scheme. In the Peu a Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. The objectives of the the present study were to find a means of carrying out the exact calculations needed to analyze the Peu a Peu fuel-loading scheme and to optimize the fuel composition, and fuel-loading scheme to achieve better burnup characteristics. The Monte Carlo method is used to perform calculations with high accuracy. Before the calculation of the whole core, the analysis for the infinite geometry was performed. The power generated per mass consumed for each combination of the uranium enrichment and packing fraction was analyzed from the parametric survey. By using the optimal value obtained, a whole-core calculation for the small 20 MWth reactor was performed and the criticality and burnup of this design was analyzed.

  8. Fast quench reactor and method

    SciTech Connect

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    2002-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  9. Fast quench reactor and method

    SciTech Connect

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    1998-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  10. Fast quench reactor and method

    SciTech Connect

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    2002-09-24

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  11. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    SciTech Connect

    Wiesenack, W.

    1996-03-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project`s data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup.

  12. Monte Carlo studies on the burnup measurement for the high temperature gas cooling reactor

    NASA Astrophysics Data System (ADS)

    Yan, Wei-Hua; Zhang, Li-Guo; Zhang, Yan; Zhang, Zhao; Xiao, Zhi-Gang

    2013-11-01

    Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanium (HPGe) gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the 137Cs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% (1σ). The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors.

  13. A fuel for sub-critical fast reactor

    NASA Astrophysics Data System (ADS)

    Moiseenko, V. E.; Chernitskiy, S. V.; Ågren, O.; Noack, K.

    2012-06-01

    Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and 238U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

  14. A fuel for sub-critical fast reactor

    SciTech Connect

    Moiseenko, V. E.; Chernitskiy, S. V.; Agren, O.; Noack, K.

    2012-06-19

    Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and {sup 238}U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

  15. Heterogeneous Transmutation Sodium Fast Reactor

    SciTech Connect

    S. E. Bays

    2007-09-01

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

  16. Stationary Liquid Fuel Fast Reactor

    SciTech Connect

    Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  17. Antineutrinos for Reactor Safeguards: Effect of Fuel Loading and Burnup on the Signal

    NASA Astrophysics Data System (ADS)

    Erickson, Anna; Bernstein, Adam; Bowden, Nathaniel

    2014-02-01

    Various types of nuclear reactor related information, including relative power level and fuel evolution parameters, can be inferred remotely using antineutrino detectors. We show that it is possible to verify assembly-level burnup using information derived from an antineutrino detector if the nominal reactor fuel loading is known. Alternatively, if the core power is measured using an independent method, for example, a thermal hydraulic element, and the nominal core behavior is known, the antineutrino detector has a capability to determine previously unknown MOX loading in the core.

  18. Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

    SciTech Connect

    G. S. Chang

    2006-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

  19. On fast reactor kinetics studies

    SciTech Connect

    Seleznev, E. F.; Belov, A. A.; Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F.

    2012-07-01

    The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

  20. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  1. Analysis of Burnup and Economic Potential of Alternative Fuel Materials in Thermal Reactors

    SciTech Connect

    Oggianu, Stella Maris; No, Hee Cheon; Kazimi, Mujid S.

    2003-09-15

    A strategy is proposed for the assessment of nuclear fuel material economic potential use in future light water reactors (LWRs). In this methodology, both the required enrichment and the fuel performance limits are considered. In order to select the best fuel candidate, the optimal burnup that produces the lowest annual fuel cost within the burnup potential for a given fuel material and smear density ratio is determined.Several nuclear materials are presented as examples of the application of the methodology proposed in this paper. The alternative fuels considered include uranium dioxide (UO{sub 2}), uranium carbide (UC), uranium nitride (UN), metallic uranium (U-Zr alloy), combined thorium and uranium oxides (ThO{sub 2}/UO{sub 2}), and combined thorium and uranium metals (U/Th). For these examples, a typical LWR lattice geometry in a zirconium-based cladding was assumed. The uncertainties in the results presented are large due to the scarcity of experimental data regarding the behavior of the considered materials at high burnups. Also, chemical compatibility issues are to be considered separately.The same methodology can be applied in the future to evaluate the economic potential of other nuclear fuel materials including different cladding designs, dispersions of ceramics into ceramics, dispersions of ceramics into metals, and also for geometries other than the traditional circular fuel pin.

  2. Method of locating a leaking fuel element in a fast breeder power reactor

    DOEpatents

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  3. HTGR Reactor Physics and Burnup Calculations Using the Serpent Monte Carlo Code

    SciTech Connect

    Leppanen, Jaakko; DeHart, Mark D

    2009-01-01

    One of the main advantages of the continuous-energy Monte Carlo method is its versatility and the capability to model any fuel or reactor configuration without major approximations. This capability becomes particularly valuable in studies involving innovative reactor designs and next-generation systems, which often lie beyond the capabilities of deterministic LWR transport codes. In this study, a conceptual prismatic HTGR fuel assembly was modeled using the Serpent Monte Carlo reactor physics burnup calculation code, under development at VTT Technical Research Centre of Finland since 2004. A new explicit particle fuel model was developed to account for the heterogeneity effects. The results are compared to other Monte Carlo and deterministic transport codes and the study also serves as a test case for the modules and methods in SCALE 6.

  4. Heterogeneous Recycling in Fast Reactors

    SciTech Connect

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  5. Fast breeder reactor protection system

    DOEpatents

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  6. Advances by the Integral Fast Reactor Program

    SciTech Connect

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs.

  7. Fast pulse nonthermal plasma reactor

    DOEpatents

    Rosocha, Louis A.

    2005-06-14

    A fast pulsed nonthermal plasma reactor includes a discharge cell and a charging assembly electrically connected thereto. The charging assembly provides plural high voltage pulses to the discharge cell. Each pulse has a rise time between one and ten nanoseconds and a duration of three to twenty nanoseconds. The pulses create nonthermal plasma discharge within the discharge cell. Accordingly, the nonthermal plasma discharge can be used to remove pollutants from gases or break the gases into smaller molecules so that they can be more efficiently combusted.

  8. Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

    SciTech Connect

    G. S. Chang

    2005-08-01

    A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data.

  9. Research Program of a Super Fast Reactor

    SciTech Connect

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki; Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki; GOTO, Shoji

    2006-07-01

    Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

  10. Small LBE-Cooled Fast Reactor for Expanding Market

    SciTech Connect

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

    2002-07-01

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

  11. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  12. High-burnup core design using minor actinide-containing metal fuel

    SciTech Connect

    Ohta, Hirokazu; Ogata, Takanari; Obara, T.

    2013-07-01

    A neutronic design study of metal fuel fast reactor (FR) cores is conducted on the basis of an innovative fuel design concept to achieve an extremely high burnup and realize an efficient fuel cycle system. Since it is expected that the burnup reactivity swing will become extremely large in an unprecedented high burnup core, minor actinides (MAs) from light water reactors (LWRs) are added to fresh fuel to improve the core internal conversion. Core neutronic analysis revealed that high burnups of about 200 MWd/kg for a small-scale core and about 300 MWd/kg for a large-scale core can be attained while suppressing the burnup reactivity swing to almost the same level as that of conventional cores with normal burnup. An actinide burnup analysis has shown that the MA consumption ratio is improved to about 60% and that the accumulated MAs originating from LWRs can be efficiently consumed by the high-burnup metal fuel FR. (authors)

  13. Analysis of fission-product effects in a Fast Mixed-Spectrum Reactor concept

    SciTech Connect

    White, J.R.; Burns, T.J.

    1980-02-01

    The Fast Mixed-Spectrum Reactor (FMSR) concept has been proposed by BNL as a means of alleviating certain nonproliferation concerns relating to civilian nuclear power. This breeder reactor concept has been tailored to operate on natural uranium feed (after initial startup), thus eliminating the need for fuel reprocessing. The fissile material required for criticality is produced, in situ, from the fertile feed material. This process requires that large burnup and fluence levels be achievable, which, in turn, necessarily implies that large fission-product inventories will exist in the reactor. It was the purpose of this study to investigate the effects of large fission-product inventories and to analyze the effect of burnup on fission-product nuclide distributions and effective cross sections. In addition, BNL requested that a representative 50-group fission-product library be generated for use in FMSR design calculations.

  14. Fission energy: The integral fast reactor

    SciTech Connect

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements.

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

    NASA Astrophysics Data System (ADS)

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

    2007-09-01

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

  16. The effects of spectral shift absorbers on the design and safety of fast spectrum space reactors

    NASA Astrophysics Data System (ADS)

    King, Jeffrey Charles

    Spectral Shift Absorbers (SSAs) are incorporated into space reactors to maintain them sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. The effect of four SSAs (samarium-149, europium-151, gadolinium-155, and gadolinium-157) on the submersion criticality, operation, and temperature reactivity feedback of the thermal spectrum reactors developed in the Systems for Nuclear Auxilary Power (SNAP) program is extensively documented. Recent work on SSAs in fast spectrum space reactors, preferred for compactness and higher powers, has focused on rhenium as the primary SSA. In addition to identifying additional SSAs, the present work investigates the effects of SSAs on the overall size and mass, temperature reactivity feedback, and operational lifetime of fast spectrum space reactors. The fast spectrum S4 reactor has a sectored Mo-14%Re solid-core, loadedwith UN fuel, cooled by He-30%Xe, and designed to avoid single point failures at a steady thermal power of 550 kWth. The addition of SSAs to the reactor core increases the fuel enrichment and decreases the size and mass of the reactor and the radiation shadow shield. SSA additions of boron-10, europium-151, gadolinium-155 and iridium result in the smallest and lightest S4 reactors. The effects of SSA additions on the operational lifetime and the temperature and burnup reactivity coefficients of the S^4 reactor are studied. An increasein fuel enrichment with SSAs markedly increases the operational lifetime by decreasing the burnup reactivity coefficient with only a slight decrease in the temperature reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt%), the temperature and burnup reactivity coefficients are the highest (-0.2709 ¢/K and -1.3470 /atom%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to -0.2649 ¢/K and -1.0230 /atom%, and

  17. A compact breed and burn fast reactor using spent nuclear fuel blanket

    SciTech Connect

    Hartanto, D.; Kim, Y.

    2012-07-01

    A long-life breed-and-burn (B and B) type fast reactor has been investigated from the neutronics points of view. The B and B reactor has the capability to breed the fissile fuels and use the bred fuel in situ in the same reactor. In this work, feasibility of a compact sodium-cooled B and B fast reactor using spent nuclear fuel as blanket material has been studied. In order to derive a compact B and B fast reactor, a tight fuel lattice and relatively large fuel pin are used to achieve high fuel volume fraction. The core is initially loaded with an LEU (Low Enriched Uranium) fuel and a metallic fuel is used in the core. The Monte Carlo depletion has been performed for the core to see the long-term behavior of the B and B reactor. Several important parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, fission power, and fast neutron fluence, are analyzed through Monte Carlo reactor analysis. Evolution of the core fuel composition is also analyzed as a function of burnup. Although the long-life small B and B fast reactor is found to be feasible from the neutronics point of view, it is characterized to have several challenging technical issues including a very high fast neutron fluence of the structural materials. (authors)

  18. Closing nuclear fuel cycle with fast reactors: problems and prospects

    SciTech Connect

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V.

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  19. In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G

    2008-04-16

    A novel in-situ gross defect verification method for light water reactor spent fuel assemblies was developed and investigated by a Monte Carlo study. This particular method is particularly effective for old pressurized water reactor spent fuel assemblies that have natural uranium in their upper fuel zones. Currently there is no method or instrument that does verification of this type of spent fuel assemblies without moving the spent fuel assemblies from their storage positions. The proposed method uses a tiny neutron detector and a detector guiding system to collect neutron signals inside PWR spent fuel assemblies through guide tubes present in PWR assemblies. The data obtained in such a manner are used for gross defect verification of spent fuel assemblies. The method uses 'calibration curves' which show the expected neutron counts inside one of the guide tubes of spent fuel assemblies as a function of fuel burn-up. By examining the measured data in the 'calibration curves', the consistency of the operator's declaration is verified.

  20. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    SciTech Connect

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; Fulwyler, James

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  1. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    DOE PAGESBeta

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; et al

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  2. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  3. Modular Code and Data System for Fast Reactor Neutronics Analyses

    SciTech Connect

    RIMPAULT, G.

    2008-06-30

    Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&D organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected in the

  4. Modular Code and Data System for Fast Reactor Neutronics Analyses

    Energy Science and Technology Software Center (ESTSC)

    2008-06-30

    Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&Dmore » organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected

  5. COUPLED FAST-THERMAL POWER BREEDER REACTOR

    DOEpatents

    Avery, R.

    1961-07-18

    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  6. Fast-acting nuclear reactor control device

    DOEpatents

    Kotlyar, Oleg M.; West, Phillip B.

    1993-01-01

    A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

  7. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    NASA Astrophysics Data System (ADS)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  8. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    SciTech Connect

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-31

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  9. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    NASA Astrophysics Data System (ADS)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

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

    SciTech Connect

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

    2015-06-17

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

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

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

    SciTech Connect

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

    2015-10-01

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

  13. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    SciTech Connect

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

  14. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    NASA Astrophysics Data System (ADS)

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Aziz, Ferhat; Permana, Sidik; Sekimoto, Hiroshi

    2014-02-01

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  15. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    SciTech Connect

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

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

    SciTech Connect

    Samuel Bays; Ayodeji Alajo

    2010-05-01

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

  17. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    SciTech Connect

    Asmolov, V.; Yegorova, L.

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

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

    DOE PAGESBeta

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

    2015-06-17

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

  19. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    NASA Astrophysics Data System (ADS)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  20. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  1. Integral data for fast reactors

    SciTech Connect

    Collins, P.J.; Poenitz, W.P.; McFarlane, H.F.

    1988-01-01

    Requirements at Argonne National Laboratory to establish the best estimates and uncertainties for LMR design parameters have lead to an extensive evaluation of the available critical experiment database. Emphasis has been put upon selection of a wide range of cores, including both benchmark, assemblies covering a range of spectra and compositions and power reactor mock-up assemblies with diverse measured parameters. The integral measurements have been revised, where necessary, using the most recent reference data and a covariance matrix constructed. A sensitivity database has been calculated, embracing all parameters, which enables quantification of the relevance of the integral data to parameters calculated with ENDF/B-V.2 cross sections.

  2. MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

    DOEpatents

    Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

    1962-06-26

    A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

  3. Nonlinear, inelastic fast reactor subassembly interaction analyses

    SciTech Connect

    Sutherland, W.H.; Bard, F.E.

    1983-01-01

    Liquid Metal Fast Breeder Reactor (LMFBR) core structural design is complicated by the trade-offs associated with keeping the subassemblies closely packed for the neutronic considerations and accommodating the volumetric changes associated with irradiation swelling. The environmental variation across the reactor core results in temperature and neutron flux gradients across the subassemblies which in turn cause the subassemblies to bow as well as dilate and grow volumetrically. These deformations in a tightly packed reactor core cause the subassemblies to interact and can potentially result in excessive withdrawal loads during the refueling operations. ABADAN, a general purpose, nonlinear, inelastic, multi-dimensional finite element structural analysis computer code, was developed for the express purpose of solving large nonlinear problems as typified by the above interaction problems. For the subassembly interaction problem ABADAN has been applied to the solution of an interacting radial row of Fast Flux Test Facility (FFTF) fuel assemblies.

  4. Burnup Predictions for Metal Fuel Tests in the Fast Flux Test Facility

    SciTech Connect

    Wootan, David W.; Nelson, Joseph V.

    2012-06-01

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The FFTF operated successfully from initial startup in 1980 through the end of the last operating cycle in March, 1992. A variety of fuel tests were irradiated in FFTF to provide performance data over a range of conditions. The MFF-3 and MFF-5 tests were U10Zr metal fuel tests with HT9 cladding. The MFF-3 and MFF-5 tests were both aggressive irradiation tests of U10Zr metal fuel pins with HT9 cladding that were prototypic of full scale LMR designs. MFF-3 was irradiated for 726 Effective Full Power Days (EFPD), starting from Cycle 10C1 (from November 1988 through March 1992), and MFF-5 was irradiated for 503 EFPD starting from Cycle 11B1 (from January 1990 through March 1992). A group of fuel pins from these two tests are undergoing post irradiation examination at the Idaho National Laboratory (INL) for the Fuel Cycle Research and Development Program (FCRD). The generation of a data package of key information on the irradiation environment and current pin detailed compositions for these tests is described. This information will be used in interpreting the results of these examinations.

  5. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    SciTech Connect

    Navarro, Jorge

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  6. Fast Spectrum Molten Salt Reactor Options

    SciTech Connect

    Gehin, Jess C; Holcomb, David Eugene; Flanagan, George F; Patton, Bruce W; Howard, Rob L; Harrison, Thomas J

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  7. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    SciTech Connect

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  8. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    SciTech Connect

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  9. Gas-Fast Reactor Fuel Fabrication

    SciTech Connect

    Randall Fielding; Mitchell Meyer; Ramprashad Prabhakaran; Jim Miller; Sean McDeavitt

    2005-11-01

    The gas-cooled fast reactor is a high temperature helium cooled Generation IV reactor concept. Operating parameters for this type of reactor are well beyond those of current fuels so a novel fuel must be developed. One fuel concept calls for UC particles dispersed throughout a SiC matrix. This study examines a hybrid reaction bonding process as a possible fabrication route for this fuel. Processing parameters are also optimized. The process combines carbon and SiC powders and a carbon yielding polymer. In order to obtain dense reaction bonded SiC samples the porosity to carbon ratio in the preform must be large enough to accommodate SiC formation from the carbon present in the sample, however too much porosity reduces mechanical integrity which leads to poor infiltration properties . The porosity must also be of a suitable size to allow silicon transport throughout the sample but keep residual silicon to a minimum.

  10. Risk Management for Sodium Fast Reactors.

    SciTech Connect

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

    2015-01-01

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

  11. Preliminary TRIGA fuel burn-up evaluation by means of Monte Carlo code and computation based on total energy released during reactor operation

    SciTech Connect

    Borio Di Tigliole, A.; Bruni, J.; Panza, F.; Alloni, D.; Cagnazzo, M.; Magrotti, G.; Manera, S.; Prata, M.; Salvini, A.; Chiesa, D.; Clemenza, M.; Pattavina, L.; Previtali, E.; Sisti, M.; Cammi, A.

    2012-07-01

    Aim of this work was to perform a rough preliminary evaluation of the burn-up of the fuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory (LENA) of the Univ. of Pavia. In order to achieve this goal a computation of the neutron flux density in each fuel element was performed by means of Monte Carlo code MCNP (Version 4C). The results of the simulations were used to calculate the effective cross sections (fission and capture) inside fuel and, at the end, to evaluate the burn-up and the uranium consumption in each fuel element. The evaluation, showed a fair agreement with the computation for fuel burn-up based on the total energy released during reactor operation. (authors)

  12. Fast critical experiment data for space reactors

    NASA Astrophysics Data System (ADS)

    Collins, P. J.; McFarlane, H. F.; Olsen, D. N.; Atkinson, C. A.; Ross, J. R.

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, about 20 fast benchmark critical experiments were calculated with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors.

  13. Actinide management with commercial fast reactors

    NASA Astrophysics Data System (ADS)

    Ohki, Shigeo

    2015-12-01

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GWey if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  14. Actinide management with commercial fast reactors

    SciTech Connect

    Ohki, Shigeo

    2015-12-31

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GW{sub e}y if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  15. Accident analysis for US fast burst reactors

    SciTech Connect

    Paternoster, R.; Flanders, M.; Kazi, H.

    1994-09-01

    In the US fast burst reactor (FBR) community there has been increasing emphasis and scrutiny on safety analysis and understanding of possible accident scenarios. This paper summarizes recent work in these areas that is going on at the different US FBR sites. At this time, all of the FBR facilities have or in the process of updating and refining their accident analyses. This effort is driven by two objectives: to obtain a more realistic scenario for emergency response procedures and contingency plans, and to determine compliance with changing regulatory standards.

  16. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    SciTech Connect

    STAN, MARIUS; HECKER, SIEGFRIED S.

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  17. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    SciTech Connect

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  18. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-01

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  19. A resting bottom sodium cooled fast reactor

    SciTech Connect

    Costes, D.

    2012-07-01

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

  20. The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

    SciTech Connect

    Blandinskiy, V. Yu.

    2014-12-15

    This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

  1. The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

    NASA Astrophysics Data System (ADS)

    Blandinskiy, V. Yu.

    2014-12-01

    This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

  2. Accident source terms for boiling water reactors with high burnup cores.

    SciTech Connect

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  3. LFR "Lead-Cooled Fast Reactor"

    SciTech Connect

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

    2006-05-11

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

  4. A CANDU-Based Fast Irradiation Reactor

    SciTech Connect

    Shatilla, Youssef

    2006-07-01

    A new steady-state fast neutron reactor is needed to satisfy the testing needs of Generation IV reactors, the Space Propulsion Program, and the Advanced Fuel Cycle Initiative. This paper presents a new concept for a CANDU-based fast irradiation reactor that is horizontal in orientation, with individual pressure tubes running the entire length of the scattering-medium tank (Calandria) filled with Lead-Bismuth-Eutectic (LBE). This approach for a test reactor will provide more flexibility in refueling, sample removal, and ability to completely re-configure the core to meet different users' requirements. Full core neutronic analysis of several fuel/coolant/geometry combinations showed a small hexagonal, LBE-cooled, U-Pu-10Zr fuel, with a core power of 100 MW{sub th} produced a fast flux (>0.1 MeV) of 1.5 x 10{sup 15} n/cm{sup 2} sec averaged over the whole length of six irradiation channels with a total testing volume of more than 77 liters. In-core breeding allowed the Pu-239 enrichment to be 15.3% which should result in core continuous operation for 180 effective full power days. Other coolants investigated included high pressure water steam and helium. An innovative shutdown/control system which consisted of the six outermost fuel channels was proven to be effective in shutting the core down when flooded with boric acid as a neutron absorber. The new shutdown/control system has the advantage of causing the minimum perturbation of the axial flux shape when the control channels are partially flooded with boric acid. This is because the acid is injected homogeneously along the control channel in contrast to regular control rods that are injected partially causing an axial perturbation in the core flux which in turn reduces safety analysis margins. The new shutdown/control system is not required to penetrate the core in a direction vertical to the fuel channels which allowed the freedom of changing core pitch as deemed necessary. A preliminary thermal hydraulic analysis

  5. The Industrial Sodium Cooled Fast Reactor

    SciTech Connect

    Samuel E. Bays; Haihua Zhao; Hongbin Zhang

    2009-04-01

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

  6. A survey of alternative once-through fast reactor core designs

    SciTech Connect

    Fei, T.; Richard, J. G.; Kersting, A. R.; Don, S. M.; Oi, C.; Driscoll, M. J.; Shwageraus, E.

    2012-07-01

    Reprocessing of Light Water Reactor (LWR) spent fuel to recover plutonium or transuranics for use in Sodium cooled Fast Reactors (SFRs) is a distant prospect in the U.S.A. This has motivated our evaluation of potentially cost-effective operation of uranium startup fast reactors (USFRs) in a once-through mode. This review goes beyond findings reported earlier based on a UC fueled MgO reflected SFR to describe a broader parametric study of options. Cores were evaluated for a variety of fuel/coolant/reflector combinations: UC/UZr/UO{sub 2}/UN;Na/Pb; MgO/SS/Zr. The challenge is achieving high burnup while minimizing enrichment and respecting both cladding fluence/dpa and reactivity lifetime limits. These parametric studies show that while UC fuel is still the leading contender, UO{sub 2} fuel and ZrH 1.7 moderated metallic fuel are also attractive if UC proves to be otherwise inadequate. Overall, these findings support the conclusion that a competitive fuel cycle cost and uranium utilization compared to LWRs is possible for SFRs operated on a once-through uranium fueled fuel cycle. In addition, eventual transition to TRU recycle mode is studied, as is a small test reactor to demonstrate key features. (authors)

  7. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    SciTech Connect

    Meriyanti; Su'ud, Zaki; Rijal, K.; Zuhair; Ferhat, A.; Sekimoto, H.

    2010-06-22

    In this study a feasibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850 deg. C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticality was obtained for this reactor.

  8. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    NASA Astrophysics Data System (ADS)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  9. Direct Deterministic Method for Neutronics Analysis and Computation of Asymptotic Burnup Distribution in a Recirculating Pebble-Bed Reactor

    SciTech Connect

    Terry, William Knox; Gougar, Hans D; Ougouag, Abderrafi Mohammed-El-Ami

    2002-07-01

    A new deterministic method has been developed for the neutronics analysis of a pebble-bed reactor (PBR). The method accounts for the flow of pebbles explicitly and couples the flow to the neutronics. The method allows modeling of once-through cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times. This new work is distinguished from older methods by the systematically semi-analytical approach it takes. In particular, whereas older methods use the finite-difference approach (or an equivalent one) for the discretization and the solution of the burnup equation, the present work integrates the relevant differential equation analytically in discrete and complementary sub-domains of the reactor. Like some of the finite-difference codes, the new method obtains the asymptotic fuel-loading pattern directly, without modeling any intermediate loading pattern. This is a significant advantage for the design and optimization of the asymptotic fuel-loading pattern. The new method is capable of modeling directly both the once-through-then-out fuel cycle and the pebble recirculating fuel cycle. Although it currently includes a finite-difference neutronics solver, the new method has been implemented into a modular code that incorporates the framework for the future coupling to an efficient solver such as a nodal method and to modern cross section preparation capabilities. In its current state, the deterministic method presented here is capable of quick and efficient design and optimization calculations for the in-core PBR fuel cycle. The method can also be used as a practical "scoping" tool. It could, for example, be applied to determine the potential of the PBR for resisting nuclear-weapons proliferation and to optimize proliferation-resistant features. However, the purpose of this paper is to show that the method itself is viable. Refinements to the code are under way, with the objective of producing a powerful reactor physics

  10. BN-800 advanced nuclear power plant with fast reactor

    SciTech Connect

    Shishkin, A.N.; Kuzavkov, N.G.; Sobolev, V.A.; Shestakov, G.V.; Bagdasarov, Yu.E.; Kochetkov, L.A.; Matveyev, V.I.; Poplavsky, V.M.

    1993-12-31

    Bn-800 reactor plant with fast reactor and sodium coolant in the primary and secondary circuits is designed for operation as part of the power units in the Yuzhno-Uralskaya nuclear power plant scheduled to be constructed in Chelyabinsk region and as part unit 4 in the Beloyarskaya nuclear power plant. Reactor operations are described.

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

    SciTech Connect

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-07-01

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

  12. Fast biomass pyrolysis with an entrained-flow reactor

    NASA Astrophysics Data System (ADS)

    Bohn, M. S.; Benham, C.

    1982-02-01

    A tubular entrained flow reactor has been used to study the effect of process control variables on fast biomass pyrolysis. In this type of reactor, finely ground biomass particles are entrained by carrier gas and transported through a reactor tube which is heated to about 900 C. Biomass particles pyrolyze as a result of heat transfer from the reactor wall yielding a gas composed primarily of carbon monoxide, carbon dioxide, hydrogen, methane, and unsaturated hydrocarbons. In this experimental program three dependent variables, percent conversion to gas, gas composition, and process heat, have been measured as a function of several process control variables. These process variables include reactor temperature, carrier gas to biomass flow ratio, reactor residence time, biomass particle size, and reactor Reynolds number. The data allow one to design and predict the performance of large scale reactors and also elucidates heat transfer mechanisms in fast biomass pyrolysis.

  13. A fast spectrum dual path flow cermet reactor

    SciTech Connect

    Anghaie, S.; Feller, G.J. ); Peery, S.D.; Parsley, R.C. )

    1993-01-15

    A cermet fueled, dual path fast reactor for space nuclear propulsion applications is conceptually designed. The reactor utilizes an outer annulus core and an inner cylindrical core with radial and axial reflector. The dual path flow minimizes the impact of power peaking near the radial reflector. Basic neutronics and core design aspects of the reactor are discussed. The dual path reactor is integrated into a 25000 lbf thrust nuclear rocket.

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

    SciTech Connect

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

    2012-05-01

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

  15. Immobilization of Fast Reactor First Cycle Raffinate

    SciTech Connect

    Langley, K. F.; Partridge, B. A.; Wise, M.

    2003-02-26

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

  16. Technology Options for a Fast Spectrum Test Reactor

    SciTech Connect

    D. M. Wachs; R. W. King; I. Y. Glagolenko; Y. Shatilla

    2006-06-01

    Idaho National Laboratory in collaboration with Argonne National Laboratory has evaluated technology options for a new fast spectrum reactor to meet the fast-spectrum irradiation requirements for the USDOE Generation IV (Gen IV) and Advanced Fuel Cycle Initiative (AFCI) programs. The US currently has no capability for irradiation testing of large volumes of fuels or materials in a fast-spectrum reactor required to support the development of Gen IV fast reactor systems or to demonstrate actinide burning, a key element of the AFCI program. The technologies evaluated and the process used to select options for a fast irradiation test reactor (FITR) for further evaluation to support these programmatic objectives are outlined in this paper.

  17. Run - Beyond - Cladding - Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

    SciTech Connect

    Batte, G.L. ); Hoffman, G.L. )

    1990-01-01

    In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab.

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

    NASA Astrophysics Data System (ADS)

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

    2016-03-01

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

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

    SciTech Connect

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

    2012-07-01

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

  20. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Satya, Octavianus Cakra; Monado, Fiber; Su'ud, Zaki; Sekimoto, Hiroshi

    2016-03-01

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on "Region-8" and "Region-10" core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  1. Composite nuclear fuel fabrication methodology for gas fast reactors

    NASA Astrophysics Data System (ADS)

    Vasudevamurthy, Gokul

    An advanced fuel form for use in Gas Fast Reactors (GFR) was investigated. Criteria for the fuel includes operation at high temperature (˜1400°C) and high burnup (˜150 MWD/MTHM) with effective retention of fission products even during transient temperatures exceeding 1600°C. The GFR fuel is expected to contain up to 20% transuranics for a closed fuel cycle. Earlier evaluations of reference fuels for the GFR have included ceramic-ceramic (cercer) dispersion type composite fuels of mixed carbide or nitride microspheres coated with SiC in a SiC matrix. Studies have indicated that ZrC is a potential replacement for SiC on account of its higher melting point, increased fission product corrosion resistance and better chemical stability. The present work investigated natural uranium carbide microspheres in a ZrC matrix instead of SiC. Known issues of minor actinide volatility during traditional fabrication procedures necessitated the investigation of still high temperature but more rapid fabrication techniques to minimize these anticipated losses. In this regard, fabrication of ZrC matrix by combustion synthesis from zirconium and graphite powders was studied. Criteria were established to obtain sufficient matrix density with UC microsphere volume fractions up to 30%. Tests involving production of microspheres by spark erosion method (similar to electrodischarge machining) showed the inability of the method to produce UC microspheres in the desired range of 300 to 1200 mum. A rotating electrode device was developed using a minimum current of 80A and rotating at speeds up to 1500 rpm to fabricate microspheres between 355 and 1200 mum. Using the ZrC process knowledge, UC electrodes were fabricated and studied for use in the rotating electrode device to produce UC microspheres. Fabrication of the cercer composite form was studied using microsphere volume fractions of 10%, 20%, and 30%. The macrostructure of the composite and individual components at various stages were

  2. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    SciTech Connect

    Seifritz, W.

    1983-11-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase.

  3. Homogeneous fast-flux isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

  4. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    SciTech Connect

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal,; Monado, Fiber; Sekimoto, Hiroshi

    2012-06-06

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  5. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Monado, Fiber; Sekimoto, Hiroshi

    2012-06-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of ith region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  6. Determination of Light Water Reactor Fuel Burnup with the Isotope Ratio Method

    SciTech Connect

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2007-11-01

    For the current project to demonstrate that isotope ratio measurements can be extended to zirconium alloys used in LWR fuel assemblies we report new analyses on irradiated samples obtained from a reactor. Zirconium alloys are used for structural elements of fuel assemblies and for the fuel element cladding. This report covers new measurements done on irradiated and unirradiated zirconium alloys, Unirradiated zircaloy samples serve as reference samples and indicate starting values or natural values for the Ti isotope ratio measured. New measurements of irradiated samples include results for 3 samples provided by AREVA. New results indicate: 1. Titanium isotope ratios were measured again in unirradiated samples to obtain reference or starting values at the same time irradiated samples were analyzed. In particular, 49Ti/48Ti ratios were indistinguishably close to values determined several months earlier and to expected natural values. 2. 49Ti/48Ti ratios were measured in 3 irradiated samples thus far, and demonstrate marked departures from natural or initial ratios, well beyond analytical uncertainty, and the ratios vary with reported fluence values. The irradiated samples appear to have significant surface contamination or radiation damage which required more time for SIMS analyses. 3. Other activated impurity elements still limit the sample size for SIMS analysis of irradiated samples. The sub-samples chosen for SIMS analysis, although smaller than optimal, were still analyzed successfully without violating the conditions of the applicable Radiological Work Permit

  7. Phenomena and Parameters Important to Burnup Credit

    SciTech Connect

    Parks, C.V.

    2001-01-10

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given.

  8. Methods for quantifying uncertainty in fast reactor analyses.

    SciTech Connect

    Fanning, T. H.; Fischer, P. F.

    2008-04-07

    Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

  9. Calculated analysis of experiments in fast neutron reactors

    SciTech Connect

    Davydov, V. K. Kalugina, K. M.; Gomin, E. A.

    2012-12-15

    In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.

  10. Calculated analysis of experiments in fast neutron reactors

    NASA Astrophysics Data System (ADS)

    Davydov, V. K.; Kalugina, K. M.; Gomin, E. A.

    2012-12-01

    In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.

  11. Simulator platform for fast reactor operation and safety technology demonstration

    SciTech Connect

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  12. 70 Group Neutron Fast Reactor Cross Section Set and 25 Group Neutron Fast Reactor Cross Section Set.

    Energy Science and Technology Software Center (ESTSC)

    1984-10-29

    Version 00 These multigroup cross sections are used in fast reactor calculations. The benchmark calculations for the 23 fast critical assemblies used in the benchmark tests of JFS-2 were performed with one-dimensional diffusion theory by using the JFS-3-J2 set.

  13. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    SciTech Connect

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  14. Fuel-Cycle of 'CANDLE' Burnup with Depleted Uranium

    SciTech Connect

    Hiroshi, Sekimoto

    2006-07-01

    A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burnup strategy can derive many merits, especially from safety point of view. The change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40 % of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50 X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The equilibrium core contains a lot of instable materials such as higher actinides and fission products, the enough amounts of which can not be obtained easily. The construction of the initial core is a difficult problem. However, by using enriched uranium substituted for actinides in the equilibrium core, we can construct the initial core whose power profile is similar to the equilibrium one and will reach the equilibrium state without any big change during transient. At present we do not have any material standing for such a high burnup. However, the CANDLE burnup can be realized by employing

  15. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    SciTech Connect

    Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel; Zhang, Yongfeng; Novascone, Stephen Rhead; Medvedev, Pavel G.

    2015-09-01

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  16. A fast shutdown system for SRS (Savannah River Site) reactors

    SciTech Connect

    Baumann, N.P.

    1990-01-01

    Power has been sharply reduced at Savannah River Site (SRS) reactors in large part to ensure that no bulk boiling occurs during hypothesized loss of coolant accidents. A fast shutdown system is essential to regain much of this lost power. Computations and experiments indicate that a He-3 injection system will serve this function. Instrumented tests of a full system are planned for early 1991 for one of the SRS reactors. 4 refs., 7 figs., 1 tab.

  17. Impact of Including Higher Actinides in Fast Reactor Transmutation Analyses

    SciTech Connect

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

    2007-09-01

    Previous fast reactor transmutation studies generally disregarded higher mass minor actinides beyond Cm-246 due to various considerations including deficiencies in nuclear cross-section data. Although omission of these higher mass actinides does not significantly impact the neutronic calculations and fuel cycle performance parameters follow-on neutron dose calculations related to fuel recycling, transportation and handling are significantly impacted. This report shows that including the minor actinides in the equilibrium fast reactor calculations will increase the predicted neutron emission by about 30%. In addition a sensitivity study was initiated by comparing the impact of different cross-section evaluation file for representing these minor actinides.

  18. Spectrophotometric Procedure for Fast Reactor Advanced Coolant Manufacture Control

    NASA Astrophysics Data System (ADS)

    Andrienko, O. S.; Egorov, N. B.; Zherin, I. I.; Indyk, D. V.

    2016-01-01

    The paper describes a spectrophotometric procedure for fast reactor advanced coolant manufacture control. The molar absorption coefficient of dimethyllead dibromide with dithizone was defined as equal to 68864 ± 795 l·mole-1·cm-1, limit of detection as equal to 0.583 · 10-6 g/ml. The spectrophotometric procedure application range was found to be equal to 37.88 - 196.3 g. of dimethyllead dibromide in the sample. The procedure was used within the framework of the development of the method of synthesis of the advanced coolant for fast reactors.

  19. Instrumentation, Monitoring and NDE for New Fast Reactors

    SciTech Connect

    Bond, Leonard J.; Doctor, Steven; Bunch, Kyle; Good, Morris; Waltar, Alan E.

    2007-07-01

    The Global Nuclear Energy Partnership (GNEP) will require the development of actinide transmutation, which can most effectively be accomplished in a fast-spectrum reactor. To achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required-- during both fabrication and operation. This paper reports parts of a knowledge capture and technology state-of-the-art assessment for fast-reactor instrumentation and controls, monitoring and diagnostics. (authors)

  20. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect

    M. K. Meyer

    2006-06-01

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

  1. Fuel development for gas-cooled fast reactors

    NASA Astrophysics Data System (ADS)

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

    2007-09-01

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

  2. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao

    2009-05-01

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

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

    SciTech Connect

    Kaito, Takeji; Ohtsuka, Satoshi; Inoue, Masaki

    2007-07-01

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

  4. Integral Fast Reactor Program. Annual progress report, FY 1992

    SciTech Connect

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1993-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  5. Detailed calculations of minor actinide transmutation in a fast reactor

    SciTech Connect

    Takeda, Toshikazu

    2015-12-31

    The transmutation of minor actinides in a fast reactor is investigated by a new method to investigate the transmutation behavior of individual minor actinides. It is found that Np-237 and Am-241 mainly contributes to the transmutation rate though the transmutation behaviors are very different.

  6. Metallic Fast Reactor Fuel Fabrication for Global Nuclear Energy Partnership

    SciTech Connect

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter

    2009-07-01

    Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt.

  7. Integral Fast Reactor Program annual progress report, FY 1991

    SciTech Connect

    Not Available

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  8. Integral Fast Reactor Program annual progress report, FY 1991

    SciTech Connect

    Not Available

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R D.

  9. Packed rod neutron shield for fast nuclear reactors

    DOEpatents

    Eck, John E.; Kasberg, Alvin H.

    1978-01-01

    A fast neutron nuclear reactor including a core and a plurality of vertically oriented neutron shield assemblies surrounding the core. Each assembly includes closely packed cylindrical rods within a polygonal metallic duct. The shield assemblies are less susceptible to thermal stresses and are less massive than solid shield assemblies, and are cooled by liquid coolant flow through interstices among the rods and duct.

  10. Integral Fast Reactor Program. Annual progress report, FY 1993

    SciTech Connect

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D.

  11. Integral Fast Reactor Program annual progress report, FY 1994

    SciTech Connect

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, J.J.

    1994-12-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R&D.

  12. Detailed calculations of minor actinide transmutation in a fast reactor

    NASA Astrophysics Data System (ADS)

    Takeda, Toshikazu

    2015-12-01

    The transmutation of minor actinides in a fast reactor is investigated by a new method to investigate the transmutation behavior of individual minor actinides. It is found that Np-237 and Am-241 mainly contributes to the transmutation rate though the transmutation behaviors are very different.

  13. Interim status report on lead-cooled fast reactor (LFR) research and development.

    SciTech Connect

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup

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

    SciTech Connect

    Samuel E. Bays

    2007-09-01

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

  15. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    SciTech Connect

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  16. Neutron initiation probability in fast burst reactor

    SciTech Connect

    Liu, X.; Du, J.; Xie, Q.; Fan, X.

    2012-07-01

    Based on the probability balance of neutron random events in multiply system, the four random process of neutron in prompt super-critical is described and then the equation of neutron initiation probability W(r,E,{Omega},t) is deduced. On the assumption of static, slightly prompt super-critical and the two factorial approximation, the formula of the average probability of 'one' neutron is derived which is the same with the result derived from the point model. The MC simulation using point model is applied in Godiva- II and CFBR-II, and the simulation result of one neutron initiation is well consistent with the theory that the initiation probability of Godiva- II inverted commas CFBR-II burst reactor are 0.00032, 0.00027 respectively on the ordinary burst operation. (authors)

  17. High Burnup Fuel Behavior Modeling

    SciTech Connect

    Jahingir, M.; Rand, R.; Stachowski, R.; Miles, B.; Kusagaya, K.

    2007-07-01

    This paper discusses the development and qualification of the PRIME03 code to address high burnup mechanisms and to improve uranium utilization in current and new reactor designs. Materials properties and behavioral models have been updated from previous thermal-mechanical codes to reflect the effects of burnup on fuel pellet thermal conductivity, Zircaloy creep, fuel pellet relocation, and fission gas release. These new models are based on results of in-pool and post irradiation examination (PIE) of commercial boiling water reactor (BWR) fuel rods at high burnup and results from international experimental programs. The new models incorporated into PRIME03 also address specific high burnup effects associated with formation of pellet rim porosity at high exposure. The PRIME03 code is qualified by comparison of predicted and measured fuel performance parameters for a large number of high, low, and moderate burnup test and commercial reactor rod. The extensive experimental qualification of the PRIME03 prediction capabilities confirms that it is a reliable best-estimate predictor of fuel rod thermal-mechanical performance over a wide range of design and operating conditions. (authors)

  18. 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.

  19. Development of fast breeder reactor fuel reprocessing technology at the Power Reactor and Nuclear Fuel Development Corporation

    SciTech Connect

    Kawata, T.; Takeda, H.; Togashi, A.; Hayashi, S. . Tokai Works); Stradley, J.G. )

    1991-01-01

    For the past two decades, a broad range of research development (R D) programs to establish fast breeder reactor (FBR) system and its associated fuel cycle technology have been pursued by the Power Reactor and Nuclear Fuel Development Corporation (PNC). Developmental activities for FBR fuel reprocessing technology have been primarily conducted at PNC Tokai Works where many important R D facilities for nuclear fuel cycle are located. These include cold and uranium tests for process equipment development in the Engineering Demonstration Facilities (EDF)-I and II, and laboratory-scale hot tests in the Chemical Processing Facility (CPF) where fuel dissolution and solvent extraction characteristics are being investigated with irradiated FBR fuel pins whose burn-up ranges up to 100,000 MWd/t. An extensive effort has also been made at EDF-III to develop advanced remote technology which enables to increase plant availability and to decrease radiation exposures to the workers in future reprocessing plants. The PNC and the United States Department of Energy (USDOE) entered into the joint collaboration in which the US shares the R Ds to support FBR fuel reprocessing program at the PNC. Several important R Ds on advanced process equipment such as a rotary dissolver and a centrifugal contactor system are in progress in a joint effort with the Oak Ridge National Laboratory (ORNL) Consolidated Fuel Reprocessing Program (CFRP). In order to facilitate hot testing on advanced processes and equipment, the design of a new engineering-scale hot test facility is now in progress aiming at the start of hot operation in late 90's. 31 refs., 2 tabs.

  20. Estimation of average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors by using the {sup 134}Cs/{sup 137}Cs ratio method

    SciTech Connect

    Endo, T.; Sato, S.; Yamamoto, A.

    2012-07-01

    Average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors is estimated, using the {sup 134}Cs/{sup 137}Cs ratio method for measured radioactivities of {sup 134}Cs and {sup 137}Cs in contaminated soils within the range of 100 km from the Fukushima Dai-ichi nuclear power plants. As a result, the measured {sup 134}Cs/{sup 137}Cs ratio from the contaminated soil is 0.996{+-}0.07 as of March 11, 2011. Based on the {sup 134}Cs/{sup 137}Cs ratio method, the estimated burnup of damaged fuels is approximately 17.2{+-}1.5 [GWd/tHM]. It is noted that the numerical results of various calculation codes (SRAC2006/PIJ, SCALE6.0/TRITON, and MVP-BURN) are almost the same evaluation values of {sup 134}Cs/ {sup 137}Cs ratio with same evaluated nuclear data library (ENDF-B/VII.0). The void fraction effect in depletion calculation has a major impact on {sup 134}Cs/{sup 137}Cs ratio compared with the differences between JENDL-4.0 and ENDF-B/VII.0. (authors)

  1. Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

    SciTech Connect

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

    1989-01-01

    The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs.

  2. Fast Breeder Reactors in Sweden: Vision and Reality.

    PubMed

    Fjaestad, Maja

    2015-01-01

    The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and '60s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy. Sweden had advanced plans for a nuclear breeder program, but canceled them in the middle of the 1970s with the rise of nuclear skepticism. The article investigates the nuclear breeder as a technological vision. The nuclear breeder reactor is an example of a technological future that did not meet its industrial expectations. But that does not change the fact that the breeder was an influential technology. Decisions about the contemporary reactors were taken with the idea that in a foreseeable future they would be replaced with the efficient breeder. The article argues that general themes in the history of the breeder reactor can deepen our understanding of the mechanisms behind technological change. PMID:26334698

  3. The IAEA international conference on fast reactors and related fuel cycles: highlights and main outcomes

    SciTech Connect

    Monti, S.; Toti, A.

    2013-07-01

    The 'International Conference on Fast Reactors and Related Fuel Cycles', which is regularly held every four years, represents the main international event dealing with fast reactors technology and related fuel cycles options. Main topics of the conference were new fast reactor concepts, design and simulation capabilities, safety of fast reactors, fast reactor fuels and innovative fuel cycles, analysis of past experience, fast reactor knowledge management. Particular emphasis was put on safety aspects, considering the current need of developing and harmonizing safety standards for fast reactors at the international level, taking also into account the lessons learned from the accident occurred at the Fukushima- Daiichi nuclear power plant in March 2011. Main advances in the several key areas of technological development were presented through 208 oral presentations during 41 technical sessions which shows the importance taken by fast reactors in the future of nuclear energy.

  4. Instrumentation, Monitoring and NDE for New Fast Reactors

    SciTech Connect

    Bond, Leonard J.; Doctor, Steven R.; Bunch, Kyle J.; Good, Morris S.; Waltar, Alan E.

    2007-07-28

    The Global Nuclear Energy Partnership (GNEP) has been proposed as a viable system in which to close the fuel cycle in a manner consistent with markedly expanding the global role of nuclear power while significantly reducing proliferation risks. A key part of this system relies on the development of actinide transmutation, which can only be effectively accomplished in a fast-spectrum reactor. The fundamental physics for fast reactors is well established. However, to achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required--during both fabrication and operation. Since the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor – II (EBR-II) reactors were operational in the USA, there have been major advances in instrumentation, not the least being the move to digital systems. Some specific capabilities have been developed outside the USA, but new or at least re-established capabilities will be required. In many cases the only available information is in reports and papers. New and improved sensors and instrumentation will be required. Advanced instrumentation has been developed for high-temperature/high-flux conditions in some cases, but most of the original researchers and manufacturers are retired or no longer in business.

  5. A Comparison of Long-Lived, Prolieration Resistant Fast Reactors

    SciTech Connect

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

    2001-09-01

    Nuclear power is expected to play a significant role in meeting future electricity needs, and in significantly reducing emissions compared to fossil-fueled power plants. However, the next generation of nuclear power plants will be expected to demonstrate significant advancements in economics, safety, waste disposal, and proliferation resistance. Many reactor types have been proposed for “Generation IV”, some of which have been fast reactors. The work discussed in here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal of the entire project is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The goal of the work presented in this paper is to investigate and compare a variety of possible fuel types, looking for optimum economics for an actinide burning, low cost of electricity, reactor design using sodium or lead-bismuth as the coolant.

  6. Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Tsai, Hanchung

    2012-08-01

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could

  7. Full-length U-xPu-10Zr (x=0, 8, 19 wt%) Fast Reactor Fuel Test in FFTF

    SciTech Connect

    D. L. Porter; H.C. Tsai

    2012-08-01

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt%) metallic fast reactor test with commercial-length (91.4 cm active fuel column length) conducted to date. With few remaining test reactors there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning of life (BOL) peak cladding temperature of the hottest pin was 608?C, cooling to 522?C at end of life (EOL). Selected fuel pins were examined non destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3 cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ~0.7 X/L axial location along the fuel column. This resulted from a lower production of rare earth fission products higher in the fuel column as well as a much smaller delta-T between fuel center and cladding, and therefore less FCCI, despite the higher cladding temperature. This behavior could

  8. A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors.

    Energy Science and Technology Software Center (ESTSC)

    1987-05-14

    Version 00 SLAROM solves the neutron integral transport equations to determine the flux distribution and spectra in a fast reactor lattice and calculates cell averaged effective cross sections. The code uses multigroup data of the type in DLC-111/JFS that use Bondarenko factors for resonance effects.

  9. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

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

    SciTech Connect

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

    2015-06-01

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

  11. Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum

    SciTech Connect

    F. Delage; J. Carmack; C. B. Lee; T. Mizuno; M. Pelletier; J. Somers

    2013-10-01

    The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxide and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.

  12. Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel

    SciTech Connect

    Eliseev, V.A.; Malisheva, I.V.; Matveev, V.I.; Egorov, A.V.; Maslov, P.A.

    2013-07-01

    With the development and research into the generation IV fast sodium reactors, great attention is paid to the enhancement of the core inherent self-protection characteristics. One of the problems dealt here is connected with the reduction of the reactivity margin so that the control rods running should not result in the core overheating and melting. In this paper we consider the possibilities of improving the core of BN-1200 with oxide fuel by a known method of introducing an axial fertile layer into the core. But unlike earlier studies this paper looks at the possibility of using such a layer not only for improving breeding, but also for reducing sodium void reactivity effect (SVRE). This proposed improvement of the BN-1200 core does not solve the problem of strong interference in control and protection system (CPS) rods of BN-1200, but they reduce significantly the reactivity margin for burn-up compensation. This helps compensate all the reactivity balances in the improved core configurations without violating constraints on SVRE value.

  13. Proposed fuel cycle for the Integral Fast Reactor

    SciTech Connect

    Burris, L.; Walters, L.C.

    1985-01-01

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

  14. Computational Neutronics Methods and Transmutation Performance Analyses for Fast Reactors

    SciTech Connect

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

    2007-03-01

    The once-through fuel cycle strategy in the United States for the past six decades has resulted in an accumulation of Light Water Reactor (LWR) Spent Nuclear Fuel (SNF). This SNF contains considerable amounts of transuranic (TRU) elements that limit the volumetric capacity of the current planned repository strategy. A possible way of maximizing the volumetric utilization of the repository is to separate the TRU from the LWR SNF through a process such as UREX+1a, and convert it into fuel for a fast-spectrum Advanced Burner Reactor (ABR). The key advantage in this scenario is the assumption that recycling of TRU in the ABR (through pyroprocessing or some other approach), along with a low capture-to-fission probability in the fast reactor’s high-energy neutron spectrum, can effectively decrease the decay heat and toxicity of the waste being sent to the repository. The decay heat and toxicity reduction can thus minimize the need for multiple repositories. This report summarizes the work performed by the fuel cycle analysis group at the Idaho National Laboratory (INL) to establish the specific technical capability for performing fast reactor fuel cycle analysis and its application to a high-priority ABR concept. The high-priority ABR conceptual design selected is a metallic-fueled, 1000 MWth SuperPRISM (S-PRISM)-based ABR with a conversion ratio of 0.5. Results from the analysis showed excellent agreement with reference values. The independent model was subsequently used to study the effects of excluding curium from the transuranic (TRU) external feed coming from the LWR SNF and recycling the curium produced by the fast reactor itself through pyroprocessing. Current studies to be published this year focus on analyzing the effects of different separation strategies as well as heterogeneous TRU target systems.

  15. Steam generator for liquid metal fast breeder reactor

    DOEpatents

    Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

    1985-01-01

    Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

  16. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  17. A Comparison of Long-Lived, Proliferation Resistant Fast Reactors

    SciTech Connect

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

    2001-09-01

    Various methods have been proposed to transmute and thus consume the current inventory of trans-uranic waste that exists in spent light-water-reactor fuel. These methods include both critical and sub-critical systems. The neutronics of metallic and nitride fuels loaded with 20-30wt% light-water-reactor plutonium plus minor actinides for use in a lead-bismuth and sodium cooled fast reactor are discussed, with an emphasis on the fuel cycle life and isotopic content. Calculations show that core life can extend beyond 20 years, and the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from 0.5 to 0.9 g/MWd.

  18. Multiple reheat helium Brayton cycles for sodium fast reactors

    SciTech Connect

    Haihua Zhao; Per F. Peterson

    2008-07-01

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

  19. How Small Can Fast-Spectrum Space Reactors Get?

    SciTech Connect

    Hatton, Steven A.; El-Genk, Mohamed S.

    2006-01-20

    Fast neutron spectrum space reactors are an appropriate choice for high thermal powers, but for low powers, they may not satisfy the excess reactivity requirement while remaining sub-critical when immersed in wet sand and flooded with seawater following a launch abort accident. This paper identifies the smallest size fast spectrum, Sectored, Compact Reactor loaded with Single UN fuel pins (SCoRe-S7), which satisfy the requirements of cold clean excess reactivity > $4.00 and remains at least $1.00 subcritical at shutdown and in submersion conditions. Results indicate that increasing the diameter of the SCoRe-S core reduces its active height and the UN fuel enrichment, but increases the Spectrum-Shift Absorber (SSA) of 157GdN additive to the fuel. All SCoRe-S cores also have a 0.1 mm thick 157Gd2O3 SSA coating on the outer surface of the reactor vessel to reduce the effect of the wet sand reflector, while the SSA fuel additive reduces the effect on the criticality of the flooded reactor caused by thermal neutron fission. The active core height decreases from 42.4 cm for the smallest SCoRe-S7 to as much as to 37.4 cm for the largest core of SCoRe-S11. For a 1.8 MWth reactor thermal power the UN fuel specific power decreases from 17.0 in the SCoRe-S7 to 11.5 Wth/kg in the -S11. The corresponding reactor total mass, including the BeO reflector, increases from 440 kg to 512 kg.

  20. Practical Combinations of Light-Water Reactors and Fast-Reactors for Future Actinide Transmutation

    SciTech Connect

    Collins, Emory D; Renier, John-Paul

    2007-01-01

    Multicycle partitioning-transmutation (P-T) studies continue to show that use of existing light-water reactors (LWRs) and new advanced light-water reactors (ALWRs) can effectively transmute transuranic (TRU) actinides, enabling initiation of full actinide recycle much earlier than waiting for the development and deployment of sufficient fast reactor (FR) capacity. The combination of initial P-T cycles using LWRs/ALWRs in parallel with economic improvements to FR usage for electricity production, and a follow-on transition period in which FRs are deployed, is a practical approach to near-term closure of the nuclear fuel cycle with full actinide recycle.

  1. Comparative breeding characteristics of fusion and fast reactors.

    PubMed

    Fortescue, P

    1977-06-17

    Expressions are developed to allow ready comparison of a hybrid fission-fusion plant and a fast breeder with respect to the number of thermal reactors that their fissile production could support, both for their feed requirements and for the new inventory needs of an expanding industry. These relations are expressed in terms of the neutron multiplication factor obtained in the fusion blanket, and the analogous quantities represented by the conversion ratios of the fast and thermal fission associated with the comparison. Results are presented graphically both for the steady state and for industries of arbitrary growth rate, and include the influence of tritium production requirements. Even a modest blanket neutron multiplication factor could enable the hybrid fusion system greatly to outperform the fast breeder on this simple basis of material balances. PMID:17831749

  2. Particle-bed gas-cooled fast reactor (PB-GCFR) design. Project final technical report (Sept 2001 - Aug 2003).

    SciTech Connect

    Taiwo, T. A.; Wei, T. Y. C.; Feldman, E. E.; Hoffman, E. A.; Fatone, M.; Holland, J. W.; Prokofiev, I. G.; Yang, W. S.; Palmiotti, G.; Hill, R. N.; Todosow, M.; Salvatores, M.; Gandini, A.

    2003-10-27

    The objective of this project is to develop a conceptual design of a particle-bed, gas-cooled fast reactor (PB-GCFR) core that meets the advanced reactor concept and enhanced proliferation-resistant goals of the US Department of Energy's NERI program. The key innovation of this project is the application of a fast neutron spectrum environment to enhance both the passive safety and transmutation characteristics of the advanced particle-bed and pebble-bed reactor designs. The PB-GCFR design is expected to produce a high-efficiency system with a low unit cost. It is anticipated that the fast neutron spectrum would permit small-sized units ({approx} 150 MWe) that can be built quickly and packaged into modular units, and whose production can be readily expanded as the demand grows. Such a system could be deployed globally. The goals of this two-year project are as follows: (1) design a reactor core that meets the future needs of the nuclear industry, by being passively safe with reduced need for engineered safety systems. This will entail an innovative core design incorporating new fuel form and type; (2) employ a proliferation-resistant fuel design and fuel cycle. This will be supported by a long-life core design that is refueled infrequently, and hence, reduces the potential for fuel diversion; (3) incorporate design features that permit use of the system as an efficient transmuter that could be employed for burning separated plutonium fuel or recycled LWR transuranic fuel, should the need arise; and (4) evaluate the fuel cycle for waste minimization and for the possibility of direct fuel disposal. The application of particle-bed fuel provides the promise of extremely high burnup and fission-product protection barriers that may permit direct disposal.

  3. Reactivity Control Schemes for Fast Spectrum Space Nuclear Reactors

    SciTech Connect

    Craft, Aaron E.; King, Jeffrey C.

    2008-01-21

    Several different reactivity control schemes are considered for future space nuclear reactor power systems. Each of these control schemes uses a combination of boron carbide absorbers and/or beryllium oxide reflectors to achieve sufficient reactivity swing to keep the reactor subcritical during launch and to provide sufficient excess reactivity to operate the reactor over its expected 7-15 year lifetime. The size and shape of the control system directly impacts the size and mass of the space reactor's reflector and shadow shield, leading to a tradeoff between reactivity swing and total system mass. This paper presents a trade study of drum, shutter, and petal control schemes based on reactivity swing and mass effects for a representative fast-spectrum, gas-cooled reactor. For each control scheme, the dimensions and composition of the core are constant, and the reflector is sized to provide $5 of cold-clean excess reactivity with each configuration in its most reactive state. The advantages and disadvantages of each configuration are discussed, along with optimization techniques and novel geometric approaches for each scheme.

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

    SciTech Connect

    Douglas L. Porter

    2011-02-01

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

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

    SciTech Connect

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

    2008-04-01

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

  6. Fast Pyrolysis of Agricultural Wastes in a Fluidized Bed Reactor

    NASA Astrophysics Data System (ADS)

    Wang, X. H.; Chen, H. P.; Yang, H. P.; Dai, X. M.; Zhang, S. H.

    Solid biomass can be converted into liquid fuel through fast pyrolysis, which is convenient to be stored and transported with potential to be used as a fossil oil substitute. In China, agricultural wastes are the main biomass materials, whose pyrolysis process has not been researched adequately compared to forestry wastes. As the representative agricultural wastes in China, peanut shell and maize stalk were involved in this paper and pine wood sawdust was considered for comparing the different pyrolysis behaviors of agricultural wastes and forestry wastes. Fast pyrolysis experiments were carried out in a bench-scale fluidized-bed reactor. The bio-oil yieldsof peanut shell and maize stalk were obviously lower than that ofpine sawdust. Compared with pine sawdust, the char yields of peanut shell and maize stalk were higher but the heating value of uncondensable gaswas lower. This means that the bio-oil cost will be higher for agricultural wastes if taking the conventional pyrolysis technique. And the characteristic and component analysis resultsof bio-oil revealed that the quality of bio-oil from agricultural wastes, especially maize stalk, was worse than that from pine wood. Therefore, it is important to take some methods to improve the quality of bio-oilfrom agricultural wastes, which should promote the exploitation of Chinese biomass resources through fast pyrolysis in afluidized bed reactor.

  7. Innovative Fast Reactors: Impact of Fuel Composition on Reactivity Coefficients

    SciTech Connect

    G. Palmiotti; M. Salvatores; M. Asswaroongruengchot

    2009-12-01

    A major challenge for future Fast Reactors could be the recycling of minor actinides (MA) in the core fuel, in order to minimize wastes and contribute to meet both the sustainability objective and the reduction of the burden on a geological disposal. Although the most outstanding issues will be found in the development and validation of the appropriate fuels, the presence of MA in the core can potentially deteriorate the core reactivity coefficients. In the present paper we will show however that there is no physical limit to the amount of MA in the core fuel, but that a careful physics analysis can indicate the most appropriate measures to reduce the MA impact on the reactivity coefficients, and in particular, for Na cooled reactors, on the Na void reactivity coefficient.

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

    SciTech Connect

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

    2009-08-01

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

  9. METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

    DOEpatents

    Koch, L.J.

    1959-01-20

    A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

  10. Current status and directions for fast reactor reprocessing

    SciTech Connect

    Burch, W.D.

    1983-01-01

    The development of fast breeder reactors (FBRs) for commercial electric power production has been under way in several countries for more than 20 years. In the United States as elsewhere, early work was centered on small reactors to prove the feasibility of concepts and later was followed by larger reactors to test engineering features and to develop fuel technology. In the early 1970s, with the perceived crisis in electrical generation expected late in this century, major efforts were mounted to plan and carry out comprehensive development programs to ensure the capability to develop and begin using this new form of nuclear power by the end of this century. This comprehensive effort included the first serious efforts directed toward the supporting fuel cycle activities. However, because of the effects of the oil price rise and resulting conservation, a slowdown of industrial growth, and cut-backs in energy needs, there has been a decline in program activities. Unlike the fuel cycle for light-water reactors (LWRs), where supply and the back-end recycle and/or waste disposal activities can largely be uncoupled, recovery and recycle of fissile materials in spent fuel must be accomplished in one or two years in a practical breeder system. 3 references.

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

    SciTech Connect

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

    1994-06-01

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

  12. The effect of coolant orificing on the core performance of a heterogeneous liquid-metal fast breeder reactor

    SciTech Connect

    Mamoru, K.; Shigehiro, A.; Yoshiaki, O.

    1983-04-01

    The effect of orificing on the core performance of a commercial-size heterogeneous liquid-metal fast breeder reactor was studied analytically. The thermal power output was flattened at beginning of life, and the coolant flow rate was chosen such that the maximum inner cladding temperature of a driver fuel and a blanket fuel was less than or equal to 620/sup 0/C at both beginning of equilibrium life (BOEL) and end of equilibrium life (EOEL). The difference between reactor outlet temperatures at BOEL and EOEL was then calculated for six core configurations: one homogeneous core configuration and five heterogeneous ones. The results showed that the core outlet temperature variation due to the change of the power profile of the radial heterogeneous core configurations is similar to that of the homogeneous one, even when a single type of orificing is used in each core zone, and it will not be necessary to use the more detailed orificing in each zone of a heterogeneous core configuration. The study concludes that for the present design, especially the thermal design, of some heterogeneous core configurations, it is feasible to control the change of the reactor outlet temperature with burnup, even when a single type of orificing is used in each core zone.

  13. Fast reactor core concepts to improve transmutation efficiency

    NASA Astrophysics Data System (ADS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-12-01

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  14. Fast reactor core concepts to improve transmutation efficiency

    SciTech Connect

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-12-31

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  15. Steam generator tubing development for commercial fast breeder reactors

    SciTech Connect

    Sessions, C.E.; Uber, C.F.

    1981-11-01

    The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs.

  16. Five Requirements for Nuclear Energy and CANDLE Fast Reactor

    NASA Astrophysics Data System (ADS)

    Sekimoto, Hiroshi

    2010-06-01

    The Center for Research into Innovative Nuclear Energy Systems (CRINES) was established in order to succeed the COE-INES mission after finishing this program in Tokyo Tech. CRINES considers nuclear energy should satisfy 5 requirements; sustainability as basic energy, solving 3 problems inherent to accidents, radioactive waste and nuclear bomb, and economical acceptance. Characteristics of CANDLE fast reactor are discussed for these requirements. It satisfies 4 requirements; sustainability and solving 3 inherent problems. For the remaining requirement for economy, a high potential to satisfy this requirement is also shown.

  17. Five Requirements for Nuclear Energy and CANDLE Fast Reactor

    SciTech Connect

    Sekimoto, Hiroshi

    2010-06-22

    The Center for Research into Innovative Nuclear Energy Systems (CRINES) was established in order to succeed the COE-INES mission after finishing this program in Tokyo Tech. CRINES considers nuclear energy should satisfy 5 requirements; sustainability as basic energy, solving 3 problems inherent to accidents, radioactive waste and nuclear bomb, and economical acceptance. Characteristics of CANDLE fast reactor are discussed for these requirements. It satisfies 4 requirements; sustainability and solving 3 inherent problems. For the remaining requirement for economy, a high potential to satisfy this requirement is also shown.

  18. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    SciTech Connect

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  19. Shape optimization of a sodium cooled fast reactor

    NASA Astrophysics Data System (ADS)

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

    2014-06-01

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

  20. The search for advanced remote technology in fast reactor reprocessing

    SciTech Connect

    Burch, W.D.; Herndon, J.N.; Stradley, J.G.

    1990-01-01

    Research and development in fast reactor reprocessing has been under way about 20 years in several countries throughout the world. During the past decade in France and the United Kingdom, active development programs have been carried out in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the EBR-II facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. Germany and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in all of these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper will focus principally on the search for improved facility concepts and better maintenance systems in the CFRP and, in turn, on how developments at ORNL have influenced the technology elsewhere.

  1. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    SciTech Connect

    A.H. Wells

    2004-11-17

    The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

  2. The development and application of an improved reactor analysis model for fast reactors

    NASA Astrophysics Data System (ADS)

    Hou, Jia

    Accuracy in neutron cross sections calculation and consistency in reactor physics are fundamental requirements in advanced nuclear reactor design and analysis. The work presented in this dissertation focuses on the development and advanced application of a reactor analysis model with updated cross section libraries that is suitable for online cross section generation for fast reactors. Research has been performed in two areas of interest in reactor physics. The first target of the research is to develop effcient modeling capacity of the 1- D lattice code MICROX-2 for its neutron spectrum calculation based on Collision Probability Method (CPM). Expanded master cross section libraries have been generated based on updated nuclear data and optimized fine-group energy structure to accommodate both thermal and fast reactor spectra as well as to comply with the need for advanced fuel cycle analysis. After verifying the new libraries, the solution methods have been reviewed and updated, including the update of interpolation scheme for resonance self-shielding factors and improvement of spatial self-shielding models for various fuel assembly geometries. The assessment of the updated lattice calculation models has shown that the prediction accuracy of lattice properties represented by the eigenvalue and reaction rate ratios is improved, especially for fast neutron spectrum lattices of which the importance of neutrons in the unresolved energy range is high. The second target of the research is to improve the accuracy of few-group nuclear cross section generation for the reactor core calculation. A 2-D pin-by-pin lattice model has been developed based on embedded CPM within the framework of the Nodal Expansion Method (NEM), which is capable of modeling the heterogeneity of the fuel assembly. Then, an online cross section generation methodology along with discontinuity factors has been developed based on Iterative Diffusion- Diffusion Methodology (IDDM), which can minimize the

  3. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    SciTech Connect

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-07-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B{sub 4}C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

  4. FFTF (Fast Flux Test Facility) reactor shutdown system reliability reevaluation

    SciTech Connect

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations.

  5. Designing a SCADA system simulator for fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

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

    SciTech Connect

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

    2011-09-01

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

  7. Limitations of eddy current testing in a fast reactor environment

    NASA Astrophysics Data System (ADS)

    Wu, Tao; Bowler, John R.

    2016-02-01

    The feasibility of using eddy current probes for detecting flaws in fast nuclear reactor structures has been investigated with the aim of detecting defects immersed in electrically conductive coolant including under liquid sodium during standby. For the inspections to be viable, there is a need to use an encapsulated sensor system that can be move into position with the aid of visualization tools. The initial objective being to locate the surface to be investigated using, for example, a combination of electromagnetic sensors and sonar. Here we focus on one feature of the task in which eddy current probe impedance variations due to interaction with the external surface of a tube are evaluated in order to monitor the probe location and orientation during inspection.

  8. Materials development for a fast breeder reactor steam generator concept

    SciTech Connect

    Sessions, C.E.; Reynolds, S.D. Jr.; Hebbar, M.A.; Lewis, J.F.; Kiefer, J.H.

    1981-11-01

    The progress achieved since 1977 in the important area of materials and processes development of fast reactor steam generator development is summarized. The two distinguishing features of the proposed Westinghouse-Tampa steam generator concept are the convoluted shell expansion joint (CSEJ) and the double-wall tubing with a third fluid leak detection capability. A 2/one quarter/ Cr-1 Mo low alloy steel will be used for all important parts of the generator including the CSEJ and the tubes. Other areas in which progress was made include tube-to-tubesheet (T/TS) welding, post-weld heat treatment (PWHT), tube expansion, and development of materials specifications for prototype and future plant materials. 8 refs.

  9. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

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

    SciTech Connect

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

    2010-09-01

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

  11. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    SciTech Connect

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  12. Reactor Physics Characterization of Transmutation Targeting Options in a Sodium Fast Reactor

    SciTech Connect

    Samuel E. Bays

    2007-04-01

    In sodium fast reactor designs, the fuel related inherent negative reactivity feedback is accomplished mainly through parasitic capture in U-238. However for an efficient minor actinide burning system, it is desirable to reduce or eliminate U-238 entirely to suppress further transuranic actinide generation. Consequently, reactivity feedback is accomplished by enhancing axial neutron streaming during a loss of coolant void situation. This is done by flattening “pancake” the active core geometry. Flattening the reactor also increases axial leakage which removes neutrons that could otherwise be used to destroy minor actinides. Therefore, it is important to tailor the neutron spectrum in the core for optimized feedback and minor actinide destruction simultaneously by using minor actinide and fission product targets.

  13. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  14. Medium Power Lead Alloy Fast Reactor Balance of Plant Options

    SciTech Connect

    Vaclav Dosta; Pavel Hejzlar; Neil E. Todreas; Jacopo Buongiorno

    2004-09-01

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature (~550°C) compared to that of light water reactors (~300°C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO2 (S-CO2) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545°C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312°C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO2 cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be the S-CO2 recompression cycle, because it

  15. Shear punch testing of candidate reactor materials after irradiation in fast reactors and spallation environments

    NASA Astrophysics Data System (ADS)

    Maloy, S. A.; Romero, T. J.; Hosemann, P.; Toloczko, M. B.; Dai, Y.

    2011-10-01

    Ferritic/martensitic steels and nickel-base superalloys are potential materials for use in spallation targets and fusion and fast reactors. To investigate the effects of irradiation on these materials, tests were performed after irradiation in the high energy proton beam at the Paul Scherrer Institute (SINQ Target Irradiation Program (STIP), 570 MeV), as well on specimens obtained from a driver duct irradiated in the Fast Flux Test Facility (FFTF). Dose accumulations were up to 18 dpa for STIP irradiations (at 147-406 °C) and up to 155 dpa in FFTF (at 383-505 °C). The helium/dpa ratios ranged from 0.2 to 80 appm/dpa. Mechanical testing was performed at 25 °C. Increases in shear yield and shear maximum stress with increasing dose mirrored the results observed from companion tensile tests.

  16. Simultaneous nuclear data target accuracy study for innovative fast reactors.

    SciTech Connect

    Aliberti, G.; Palmiotti, G.; Salvatores, M.; Nuclear Engineering Division; INL; CEA Cadarache

    2007-01-01

    The present paper summarizes the major outcomes of a study conducted within a Nuclear Energy Agency Working Party on Evaluation Cooperation (NEA WPEC) initiative aiming to investigate data needs for future innovative nuclear systems, to quantify them and to propose a strategy to meet them. Within the NEA WPEC Subgroup 26 an uncertainty assessment has been carried out using covariance data recently processed by joint efforts of several US and European Labs. In general, the uncertainty analysis shows that for the wide selection of fast reactor concepts considered, the present integral parameters uncertainties resulting from the assumed uncertainties on nuclear data are probably acceptable in the early phases of design feasibility studies. However, in the successive phase of preliminary conceptual designs and in later design phases of selected reactor and fuel cycle concepts, there will be the need for improved data and methods, in order to reduce margins, both for economic and safety reasons. It is then important to define as soon as possible priority issues, i.e. which are the nuclear data (isotope, reaction type, energy range) that need improvement, in order to quantify target accuracies and to select a strategy to meet the requirements needed (e.g. by some selected new differential measurements and by the use of integral experiments). In this context one should account for the wide range of high accuracy integral experiments already performed and available in national or, better, international data basis, in order to indicate new integral experiments that will be needed to account for new requirements due to innovative design features, and to provide the necessary full integral data base to be used for validation of the design simulation tools.

  17. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    SciTech Connect

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  18. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    SciTech Connect

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma; Al Rashdan, Ahmad; Tsvetkov, Pavel Valeryevich; Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  19. Role of fast reactor and its cycle to reduce nuclear waste burden

    SciTech Connect

    Arie, Kazuo; Oomori, Takashi; Okita, Takeshi; Kawashima, Masatoshi; Kotake, Shoji; Fuji-ie, Yoichi

    2013-07-01

    The role of the metal fuel fast reactor with recycling of actinides and the five long-lived fission products based on the concept of the Self-Consistent Nuclear Energy System has been examined by evaluating the reduction of nuclear wastes during the transition period to this reactor system. The evaluation was done in comparison to an LWR once-through case and a conventional actinide recycling oxide fast reactor. As a result, it is quantitatively clarified that a metal fuel fast reactor with actinide and the five long-lived fission products (I{sup 129}, Tc{sup 99}, Zr{sup 93}, Cs{sup 135} and Sn{sup 126}) recycling could play a significant role in reducing the nuclear waste burden including the current LWR wastes. This can be achieved by using a fast neutron spectrum reactor enhanced with metal fuel that brings high capability as a 'waste burner'. (authors)

  20. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    SciTech Connect

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-09-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850ºC at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05).

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

    SciTech Connect

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

    2012-05-01

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

  2. One pass core design of a super fast reactor

    SciTech Connect

    Liu, Qingjie; Oka, Yoshiaki

    2013-07-01

    One pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern can significantly simplify the core concept. Upper core structure, coolant flow scheme as well as refueling procedure are as simple as in PWR. In one pass core design, supercritical-pressure water is at approximately 25.0 MPa and enters the core at 280 C. degrees and is heated up in one through flow pattern upwardly to the average outlet temperature of 500 C. degrees. Great density change in vertical direction can cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO{sub 2} fuel blanket assemblies also introduces great power increase during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at the beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of the blanket assembly. This can help to mitigate the power change in blanket due to Pu accumulation and to increase the outlet temperature of the blanket during cycle. Neutron transport and thermohydraulics coupled calculation shows that this design can satisfy the requirement in the Super FR principle for both 500 C. degrees outlet temperature and negative coolant void reactivity. (authors)

  3. Delayed neutron signal characterization in a fast reactor

    SciTech Connect

    Gross, K.C.; Strain, R.V.

    1980-01-01

    Experimental and analytical techniques have been developed for delayed neutron (DN) signal analysis and characterization that can provide diagnostic information to augment data from cover-gas analyses in the detection and identification of breached elements in an LMFBR. Eleven flow reduction tests have been run in EBR-II to provide base data support for predicting DN signal characteristics during exposed fuel operation. Results from the tests demonstrate the feasibility and practicability of response-analysis techniques for determining the transit time, T/sub tr/, for DN emitters traveling from the core to the detector, and the isotopic holdup time, T/sub h/, of DN precursors in the fuel element. T/sub tr/ has been found to vary with the relative grid location of the DN source, and T/sub h/ is affected by the form of fuel exposed to the coolant as well as the condition of the breach site. These parameters are incorporated into a mathematical formulism that enables one to compute for any exposed-fuel test an equivalent recoil area. This concept provides a basis for comparison of different run-beyond-cladding-breach tests in fast reactors.

  4. Cermet fuel for fast reactor - Fabrication and characterization

    NASA Astrophysics Data System (ADS)

    Mishra, Sudhir; Kutty, P. S.; Kutty, T. R. G.; Das, Shantanu; Dey, G. K.; Kumar, Arun

    2013-11-01

    (U, Pu)O2 ceramic fuel is the well-established fuel for the fast reactors and (U, Pu, Zr) metallic fuel is the future fuel. Both the fuels have their own merits and demerits. Optimal solution may lie in opting for a fuel which combines the favorable features of both fuel systems. The choice may be the use of cermet fuel which can be either (U, PuO2) or (Enriched U, UO2). In the present study, attempt has been made to fabricate (Natural U, UO2) cermet fuel by powder metallurgy route. Characterization of the fuel has been carried out using dilatometer, differential thermal analyzer, X-ray diffractometer, and Scanning Electron Microscope. The results show a high solidus temperature, high thermal expansion, presence of porosities, etc. in the fuel. The thermal conductivity of the fuel has also been measured. X-ray diffraction study on the fuel compact reveals presence of α U and UO2 phases in the matrix of the fuel.

  5. Alloys for a liquid metal fast breeder reactor

    DOEpatents

    Rowcliffe, Arthur F.; Bleiberg, Melvin L.; Diamond, Sidney; Bajaj, Ram

    1979-01-01

    An essentially gamma-prime precipitation-hardened iron-chromium-nickel alloy has been designed with emphasis on minimum nickel and chromium contents to reduce the swelling tendencies of these alloys when used in liquid metal fast breeder reactors. The precipitation-hardening components have been designed for phase stability and such residual elements as silicon and boron, also have been selected to minimize swelling. Using the properties of these alloys in one design would result in an increased breeding ratio over 20% cold worked stainless steel, a reference material, of 1.239 to 1.310 and a reduced doubling time from 15.8 to 11.4 years. The gross stoichiometry of the alloying composition comprises from about 0.04% to about 0.06% carbon, from about 0.05% to about 1.0% silicon, up to about 0.1% zirconium, up to about 0.5% vanadium, from about 24% to about 31% nickel, from 8% to about 11% chromium, from about 1.7% to about 3.5% titanium, from about 1.0% to about 1.8% aluminum, from about 0.9% to about 3.7% molybdenum, from about 0.04% to about 0.8% boron, and the balance iron with incidental impurities.

  6. Selection of materials for sodium fast reactor steam generators

    SciTech Connect

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

    2012-07-01

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

  7. Ultrasonic decontamination of prototype fast breeder reactor fuel pins.

    PubMed

    Kumar, Aniruddha; Bhatt, R B; Behere, P G; Afzal, Mohd

    2014-04-01

    Fuel pin decontamination is the process of removing particulates of radioactive material from its exterior surface. It is an important process step in nuclear fuel fabrication. It assumes more significance with plutonium bearing fuel known to be highly radio-toxic owing to its relatively longer biological half life and shorter radiological half life. Release of even minute quantity of plutonium oxide powder in the atmosphere during its handling can cause alarming air borne activity and may pose a severe health hazard to personnel working in the vicinity. Decontamination of fuel pins post pellet loading operation is thus mandatory before they are removed from the glove box for further processing and assembly. This paper describes the setting up of ultrasonic decontamination process, installed inside a custom built fume-hood in the production line, comprising of a cleaning tank with transducers, heaters, pin handling device and water filtration system and its application in cleaning of fuel pins for prototype fast breeder reactor. The cleaning process yielded a typical decontamination efficiency of more than 99%. PMID:24405906

  8. Reactivity and isotopic composition of spent PWR (pressurized-water-reactor) fuel as a function of initial enrichment, burnup, and cooling time

    SciTech Connect

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub infinity/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub infinity/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub infinity/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs.

  9. 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.

  10. Fast Flux Test Facility Reactor Vessel Removal Study

    SciTech Connect

    BOWMAN, B.R.

    2002-10-23

    This study assesses the feasibility of removing the FFTF reactor vessel from its current location in the reactor cavity inside the Containment vessel to a transporter for relocation to a burial pit in the 200 Area.