Benchmarking of calculation schemes in APOLLO2 and COBAYA3 for WER lattices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zheleva, N.; Ivanov, P.; Todorova, G.

This paper presents solutions of the NURISP WER lattice benchmark using APOLLO2, TRIPOLI4 and COBAYA3 pin-by-pin. The main objective is to validate MOC based calculation schemes for pin-by-pin cross-section generation with APOLLO2 against TRIPOLI4 reference results. A specific objective is to test the APOLLO2 generated cross-sections and interface discontinuity factors in COBAYA3 pin-by-pin calculations with unstructured mesh. The VVER-1000 core consists of large hexagonal assemblies with 2 mm inter-assembly water gaps which require the use of unstructured meshes in the pin-by-pin core simulators. The considered 2D benchmark problems include 19-pin clusters, fuel assemblies and 7-assembly clusters. APOLLO2 calculation schemes withmore » the step characteristic method (MOC) and the higher-order Linear Surface MOC have been tested. The comparison of APOLLO2 vs. TRIPOLI4 results shows a very close agreement. The 3D lattice solver in COBAYA3 uses transport corrected multi-group diffusion approximation with interface discontinuity factors of Generalized Equivalence Theory (GET) or Black Box Homogenization (BBH) type. The COBAYA3 pin-by-pin results in 2, 4 and 8 energy groups are close to the reference solutions when using side-dependent interface discontinuity factors. (authors)« less

Benchmark tests of JENDL-3.2 for thermal and fast reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takano, Hideki; Akie, Hiroshi; Kikuchi, Yasuyuki

1994-12-31

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k{sub eff} and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k{sub eff} reactivity worths of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas; Burns, Joseph R.

The aftermath of the Tōhoku earthquake and the Fukushima accident has led to a global push to improve the safety of existing light water reactors. A key component of this initiative is the development of nuclear fuel and cladding materials with potentially enhanced accident tolerance, also known as accident-tolerant fuels (ATF). These materials are intended to improve core fuel and cladding integrity under beyond design basis accident conditions while maintaining or enhancing reactor performance and safety characteristics during normal operation. To complement research that has already been carried out to characterize ATF neutronics, the present study provides an initial investigationmore » of the sensitivity and uncertainty of ATF systems responses to nuclear cross section data. ATF concepts incorporate novel materials, including SiC and FeCrAl cladding and high density uranium silicide composite fuels, in turn introducing new cross section sensitivities and uncertainties which may behave differently from traditional fuel and cladding materials. In this paper, we conducted sensitivity and uncertainty analysis using the TSUNAMI-2D sequence of SCALE with infinite lattice models of ATF assemblies. Of all the ATF materials considered, it is found that radiative capture in 56Fe in FeCrAl cladding is the most significant contributor to eigenvalue uncertainty. 56Fe yields significant potential eigenvalue uncertainty associated with its radiative capture cross section; this is by far the largest ATF-specific uncertainty found in these cases, exceeding even those of uranium. We found that while significant new sensitivities indeed arise, the general sensitivity behavior of ATF assemblies does not markedly differ from traditional UO2/zirconium-based fuel/cladding systems, especially with regard to uncertainties associated with uranium. We assessed the similarity of the IPEN/MB-01 reactor benchmark model to application models with FeCrAl cladding. We used TSUNAMI-IP to calculate similarity indices of the application model and IPEN/MB-01 reactor benchmark model. This benchmark was selected for its use of SS304 as a cladding and structural material, with significant 56Fe content. The similarity indices suggest that while many differences in reactor physics arise from differences in design, sensitivity to and behavior of 56Fe absorption is comparable between systems, thus indicating the potential for this benchmark to reduce uncertainties in 56Fe radiative capture cross sections.« less

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

NASA Astrophysics Data System (ADS)

Kahler, A. C.; MacFarlane, R. E.; Mosteller, R. D.; Kiedrowski, B. C.; Frankle, S. C.; Chadwick, M. B.; McKnight, R. D.; Lell, R. M.; Palmiotti, G.; Hiruta, H.; Herman, M.; Arcilla, R.; Mughabghab, S. F.; Sublet, J. C.; Trkov, A.; Trumbull, T. H.; Dunn, M.

2011-12-01

The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., "ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data," Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235U and 239Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as 236U, 238,242Pu and 241,243Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for 233U fueled systems as a function of Above-Thermal Fission Fraction remain. The comprehensive nature of this critical benchmark suite and the generally accurate calculated eigenvalues obtained with ENDF/B-VII.1 neutron cross sections support the conclusion that this is the most accurate general purpose ENDF/B cross section library yet released to the technical community.

Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

NASA Technical Reports Server (NTRS)

Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

1971-01-01

A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7 cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The experimental program consisted basically of measuring the differential neutron spectra and the changes in critical mass that accompanied the stepwise addition of (Li-7)3N, Hf, Ta, and W to a basic core fueled with U metal in a pin-type Ta honeycomb structure. In addition, experimental results were obtained on power distributions, control characteristics, neutron lifetime, and reactivity worths of numerous absorber, structural, and scattering materials.

Creating NDA working standards through high-fidelity spent fuel modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E; Gauld, Ian C; Romano, Catherine E

2012-01-01

The Next Generation Safeguards Initiative (NGSI) is developing advanced non-destructive assay (NDA) techniques for spent nuclear fuel assemblies to advance the state-of-the-art in safeguards measurements. These measurements aim beyond the capabilities of existing methods to include the evaluation of plutonium and fissile material inventory, independent of operator declarations. Testing and evaluation of advanced NDA performance will require reference assemblies with well-characterized compositions to serve as working standards against which the NDA methods can be benchmarked and for uncertainty quantification. To support the development of standards for the NGSI spent fuel NDA project, high-fidelity modeling of irradiated fuel assemblies is beingmore » performed to characterize fuel compositions and radiation emission data. The assembly depletion simulations apply detailed operating history information and core simulation data as it is available to perform high fidelity axial and pin-by-pin fuel characterization for more than 1600 nuclides. The resulting pin-by-pin isotopic inventories are used to optimize the NDA measurements and provide information necessary to unfold and interpret the measurement data, e.g., passive gamma emitters, neutron emitters, neutron absorbers, and fissile content. A key requirement of this study is the analysis of uncertainties associated with the calculated compositions and signatures for the standard assemblies; uncertainties introduced by the calculation methods, nuclear data, and operating information. An integral part of this assessment involves the application of experimental data from destructive radiochemical assay to assess the uncertainty and bias in computed inventories, the impact of parameters such as assembly burnup gradients and burnable poisons, and the influence of neighboring assemblies on periphery rods. This paper will present the results of high fidelity assembly depletion modeling and uncertainty analysis from independent calculations performed using SCALE and MCNP. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.« less

ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lell, R. M.; McKnight, R. D.; Tsiboulia, A.

2010-09-30

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physicsmore » benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specificationsa and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 11 began in early January 1958, and the Assembly 11 program ended in late January 1958. The core consisted of highly enriched uranium (HEU) plates and depleted uranium plates loaded into stainless steel drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, six columns of 0.125 in.-wide (3.175 mm) depleted uranium plates and one column of 1.0 in.-wide (25.4 mm) depleted uranium plates. The length of each column was 10 in. (254.0 mm) in each half of the core. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the depleted uranium radial blanket was approximately 14 in. (355.6 mm), and the length of the radial blanket in each half of the matrix was 22 in. (558.8 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/11, the reference critical configuration was loading 10 which was critical on January 21, 1958. Subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/11 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/11 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertainty far beyond that of the original experiment. Such a transformation is described in Section 3. It was obtained using a pair of continuous-energy Monte Carlo calculations. First, the critical configuration was modeled in full detail - every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from the detailed as-built model were used to construct a homogeneous, two-dimensional (RZ) model of ZPR-3/11 that conserved the mass of each nuclide and volume of each region. The simple cylindrical model is the criticality-safety benchmark model. The difference in the calculated k{sub eff} values between the as-built three-dimensional model and the homogeneous two-dimensional benchmark model was used to adjust the measured excess reactivity of ZPR-3/11 loading 10 to obtain the k{sub eff} for the benchmark model.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marck, Steven C. van der, E-mail: vandermarck@nrg.eu

Recent releases of three major world nuclear reaction data libraries, ENDF/B-VII.1, JENDL-4.0, and JEFF-3.1.1, have been tested extensively using benchmark calculations. The calculations were performed with the latest release of the continuous energy Monte Carlo neutronics code MCNP, i.e. MCNP6. Three types of benchmarks were used, viz. criticality safety benchmarks, (fusion) shielding benchmarks, and reference systems for which the effective delayed neutron fraction is reported. For criticality safety, more than 2000 benchmarks from the International Handbook of Criticality Safety Benchmark Experiments were used. Benchmarks from all categories were used, ranging from low-enriched uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), tomore » mixed uranium-plutonium, metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments at Lawrence Livermore National Laboratory (for {sup 6}Li, {sup 7}Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D2O, H2O, concrete, polyethylene and teflon). The new functionality in MCNP6 to calculate the effective delayed neutron fraction was tested by comparison with more than thirty measurements in widely varying systems. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra. The performance of the three libraries, in combination with MCNP6, is shown to be good. The results for the LEU-COMP-THERM category are on average very close to the benchmark value. Also for most other categories the results are satisfactory. Deviations from the benchmark values do occur in certain benchmark series, or in isolated cases within benchmark series. Such instances can often be related to nuclear data for specific non-fissile elements, such as C, Fe, or Gd. Indications are that the intermediate and mixed spectrum cases are less well described. The results for the shielding benchmarks are generally good, with very similar results for the three libraries in the majority of cases. Nevertheless there are, in certain cases, strong deviations between calculated and benchmark values, such as for Co and Mg. Also, the results show discrepancies at certain energies or angles for e.g. C, N, O, Mo, and W. The functionality of MCNP6 to calculate the effective delayed neutron fraction yields very good results for all three libraries.« less

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krass, A.W.

2005-12-19

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

Dismantlement of the TSF-SNAP Reactor Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peretz, Fred J

2009-01-01

This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

NASA Astrophysics Data System (ADS)

Burdo, James S.

This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the difference was less than 1.00, allowing for the existence of the difference within the margin of error. The second was whether the difference between the two values was big enough to prevent their error bars from overlapping. Error analysis was performed both using a one second count and pseudo-Maxwell statistics for a projected 60 second count, giving four criteria for detection. The number of guide tubes meeting these criteria was compared and graphed for each case. Further analysis at extremes of high and low enrichment and long and short burnup times was done using data from assemblies at the Beaver Valley 1 and 2 PWR. In all neutron flux cases, at least two guide tube locations meet all the criteria for detection of pin diversion. At least one location in almost all of the gamma flux cases does. These results show that placing detectors in the empty guide tubes of spent fuel bundles to identify possible pin diversion is feasible.

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

NASA Astrophysics Data System (ADS)

Susilo, J.; Suparlina, L.; Deswandri; Sunaryo, G. R.

2018-02-01

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE PAGES

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.; ...

2017-04-28

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation

DOE PAGES

Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.; ...

2016-01-01

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via highmore » performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.« less

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bardet, Philippe; Ricciardi, Guillaume

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWRmore » fuel bundle behavior during seismic transients.« less

HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.

2014-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via highmore » performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.« less

Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bowman, S.M.

1995-01-01

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies inmore » the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The other two benchmark critical calculations were the beginning-of-cycle (BOC) startup at both hot, zero-power (HZP) and HFP critical conditions. These latter calculations were used to check for consistency in the calculated results for different burnups and downtimes. The k{sub eff} results were in the range of 1.00014 to 1.00259 with a standard deviation of less than 0.001.« less

Benchmarking of HEU Mental Annuli Critical Assemblies with Internally Reflected Graphite Cylinder

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xiaobo, Liu; Bess, John D.; Marshall, Margaret A.

Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches) metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00055, 0.00055 and 0.00055 respectively, and biases to the detailed benchmark models which are -0.00179, -0.00189 and -0.00114 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified model. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF VII.1 agree well tomore » the benchmark experimental results with a difference of less than 0.2%. These are acceptable benchmark experiments for inclusion in the ICSBEP Handbook.« less

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

NASA Technical Reports Server (NTRS)

Klann, P. G.; Lantz, E.

1973-01-01

A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./ Russian Progress Report for Fiscal Year 1997, Volume 4, Part 8 - Neutron Poison Plates in Assemblies Containing Homogeneous Mixtures of Polystyrene-Moderated Plutonium and Uranium Oxides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yavuz, M.

1999-05-01

In the 1970s at the Battelle Pacific Northwest Laboratory (PNL), a series of critical experiments using a remotely operated Split-Table Machine was performed with homogeneous mixtures of (Pu-U)O{sub 2}-polystyrene fuels in the form of square compacts having different heights. The experiments determined the critical geometric configurations of MOX fuel assemblies with and without neutron poison plates. With respect to PuO{sub 2} content and moderation [H/(Pu+U)atomic] ratio (MR), two different homogeneous (Pu-U) O{sub 2}-polystyrene mixtures were considered: Mixture (1) 14.62 wt% PuO{sub 2} with 30.6 MR, and Mixture (2) 30.3 wt% PuO{sub 2} with 2.8 MR. In all mixtures, the uraniummore » was depleted to about O.151 wt% U{sup 235}. Assemblies contained copper, copper-cadmium or aluminum neutron poison plates having thicknesses up to {approximately}2.5 cm. This evaluation contains 22 experiments for Mixture 1, and 10 for Mixture 2 compacts. For Mixture 1, there are 10 configurations with copper plates, 6 with aluminum, and 5 with copper-cadmium. One experiment contained no poison plate. For Mixture 2 compacts, there are 3 configurations with copper, 3 with aluminum, and 3 with copper-cadmium poison plates. One experiment contained no poison plate.« less

A Full-Core Resonance Self-Shielding Method Using a Continuous-Energy Quasi–One-Dimensional Slowing-Down Solution that Accounts for Temperature-Dependent Fuel Subregions and Resonance Interference

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Yuxuan; Martin, William; Williams, Mark

In this paper, a correction-based resonance self-shielding method is developed that allows annular subdivision of the fuel rod. The method performs the conventional iteration of the embedded self-shielding method (ESSM) without subdivision of the fuel to capture the interpin shielding effect. The resultant self-shielded cross sections are modified by correction factors incorporating the intrapin effects of radial variation of the shielded cross section, radial temperature distribution, and resonance interference. A quasi–one-dimensional slowing-down equation is developed to calculate such correction factors. The method is implemented in the DeCART code and compared with the conventional ESSM and subgroup method with benchmark MCNPmore » results. The new method yields substantially improved results for both spatially dependent reaction rates and eigenvalues for typical pressurized water reactor pin cell cases with uniform and nonuniform fuel temperature profiles. Finally, the new method is also proved effective in treating assembly heterogeneity and complex material composition such as mixed oxide fuel, where resonance interference is much more intense.« less

Validation of updated neutronic calculation models proposed for Atucha-II PHWR. Part I: Benchmark comparisons of WIMS-D5 and DRAGON cell and control rod parameters with MCNP5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mollerach, R.; Leszczynski, F.; Fink, J.

2006-07-01

In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure-vessel design with 451 vertical coolant channels, and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update calculation methods and models (cell, supercell and reactor) was recently carried out coveringmore » cell, supercell (control rod) and core calculations. As a validation of the new models some benchmark comparisons were done with Monte Carlo calculations with MCNP5. This paper presents comparisons of cell and supercell benchmark problems based on a slightly idealized model of the Atucha-I core obtained with the WIMS-D5 and DRAGON codes with MCNP5 results. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, and more symmetric than Atucha-II Cell parameters compared include cell k-infinity, relative power levels of the different rings of fuel rods, and some two-group macroscopic cross sections. Supercell comparisons include supercell k-infinity changes due to the control rods (tubes) of steel and hafnium. (authors)« less

Safety Criticality Standards Using the French CRISTAL Code Package: Application to the AREVA NP UO{sub 2} Fuel Fabrication Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doucet, M.; Durant Terrasson, L.; Mouton, J.

2006-07-01

Criticality safety evaluations implement requirements to proof of sufficient sub critical margins outside of the reactor environment for example in fuel fabrication plants. Basic criticality data (i.e., criticality standards) are used in the determination of sub critical margins for all processes involving plutonium or enriched uranium. There are several criticality international standards, e.g., ARH-600, which is one the US nuclear industry relies on. The French Nuclear Safety Authority (DGSNR and its advising body IRSN) has requested AREVA NP to review the criticality standards used for the evaluation of its Low Enriched Uranium fuel fabrication plants with CRISTAL V0, the recentlymore » updated French criticality evaluation package. Criticality safety is a concern for every phase of the fabrication process including UF{sub 6} cylinder storage, UF{sub 6}-UO{sub 2} conversion, powder storage, pelletizing, rod loading, assembly fabrication, and assembly transportation. Until 2003, the accepted criticality standards were based on the French CEA work performed in the late seventies with the APOLLO1 cell/assembly computer code. APOLLO1 is a spectral code, used for evaluating the basic characteristics of fuel assemblies for reactor physics applications, which has been enhanced to perform criticality safety calculations. Throughout the years, CRISTAL, starting with APOLLO1 and MORET 3 (a 3D Monte Carlo code), has been improved to account for the growth of its qualification database and for increasing user requirements. Today, CRISTAL V0 is an up-to-date computational tool incorporating a modern basic microscopic cross section set based on JEF2.2 and the comprehensive APOLLO2 and MORET 4 codes. APOLLO2 is well suited for criticality standards calculations as it includes a sophisticated self shielding approach, a P{sub ij} flux determination, and a 1D transport (S{sub n}) process. CRISTAL V0 is the result of more than five years of development work focusing on theoretical approaches and the implementation of user-friendly graphical interfaces. Due to its comprehensive physical simulation and thanks to its broad qualification database with more than a thousand benchmark/calculation comparisons, CRISTAL V0 provides outstanding and reliable accuracy for criticality evaluations for configurations covering the entire fuel cycle (i.e. from enrichment, pellet/assembly fabrication, transportation, to fuel reprocessing). After a brief description of the calculation scheme and the physics algorithms used in this code package, results for the various fissile media encountered in a UO{sub 2} fuel fabrication plant will be detailed and discussed. (authors)« less

Sensitivity Analysis of OECD Benchmark Tests in BISON

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swiler, Laura Painton; Gamble, Kyle; Schmidt, Rodney C.

2015-09-01

This report summarizes a NEAMS (Nuclear Energy Advanced Modeling and Simulation) project focused on sensitivity analysis of a fuels performance benchmark problem. The benchmark problem was defined by the Uncertainty Analysis in Modeling working group of the Nuclear Science Committee, part of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD ). The benchmark problem involv ed steady - state behavior of a fuel pin in a Pressurized Water Reactor (PWR). The problem was created in the BISON Fuels Performance code. Dakota was used to generate and analyze 300 samples of 17 input parameters defining coremore » boundary conditions, manuf acturing tolerances , and fuel properties. There were 24 responses of interest, including fuel centerline temperatures at a variety of locations and burnup levels, fission gas released, axial elongation of the fuel pin, etc. Pearson and Spearman correlatio n coefficients and Sobol' variance - based indices were used to perform the sensitivity analysis. This report summarizes the process and presents results from this study.« less

Selectivity of Direct Methanol Fuel Cell Membranes.

PubMed

Aricò, Antonino S; Sebastian, David; Schuster, Michael; Bauer, Bernd; D'Urso, Claudia; Lufrano, Francesco; Baglio, Vincenzo

2015-11-24

Sulfonic acid-functionalized polymer electrolyte membranes alternative to Nafion(®) were developed. These were hydrocarbon systems, such as blend sulfonated polyetheretherketone (s-PEEK), new generation perfluorosulfonic acid (PFSA) systems, and composite zirconium phosphate-PFSA polymers. The membranes varied in terms of composition, equivalent weight, thickness, and filler and were investigated with regard to their methanol permeation characteristics and proton conductivity for application in direct methanol fuel cells. The behavior of the membrane electrode assemblies (MEA) was investigated in fuel cell with the aim to individuate a correlation between membrane characteristics and their performance in a direct methanol fuel cell (DMFC). The power density of the DMFC at 60 °C increased according to a square root-like function of the membrane selectivity. This was defined as the reciprocal of the product between area specific resistance and crossover. The power density achieved at 60 °C for the most promising s-PEEK-based membrane-electrode assembly (MEA) was higher than the benchmark Nafion(®) 115-based MEA (77 mW·cm(-2) vs. 64 mW·cm(-2)). This result was due to a lower methanol crossover (47 mA·cm(-2) equivalent current density for s-PEEK vs. 120 mA·cm(-2) for Nafion(®) 115 at 60 °C as recorded at OCV with 2 M methanol) and a suitable area specific resistance (0.15 Ohm cm² for s-PEEK vs. 0.22 Ohm cm² for Nafion(®) 115).

Selectivity of Direct Methanol Fuel Cell Membranes

PubMed Central

Aricò, Antonino S.; Sebastian, David; Schuster, Michael; Bauer, Bernd; D’Urso, Claudia; Lufrano, Francesco; Baglio, Vincenzo

2015-01-01

Sulfonic acid-functionalized polymer electrolyte membranes alternative to Nafion® were developed. These were hydrocarbon systems, such as blend sulfonated polyetheretherketone (s-PEEK), new generation perfluorosulfonic acid (PFSA) systems, and composite zirconium phosphate–PFSA polymers. The membranes varied in terms of composition, equivalent weight, thickness, and filler and were investigated with regard to their methanol permeation characteristics and proton conductivity for application in direct methanol fuel cells. The behavior of the membrane electrode assemblies (MEA) was investigated in fuel cell with the aim to individuate a correlation between membrane characteristics and their performance in a direct methanol fuel cell (DMFC). The power density of the DMFC at 60 °C increased according to a square root-like function of the membrane selectivity. This was defined as the reciprocal of the product between area specific resistance and crossover. The power density achieved at 60 °C for the most promising s-PEEK-based membrane-electrode assembly (MEA) was higher than the benchmark Nafion® 115-based MEA (77 mW·cm−2 vs. 64 mW·cm−2). This result was due to a lower methanol crossover (47 mA·cm−2 equivalent current density for s-PEEK vs. 120 mA·cm−2 for Nafion® 115 at 60 °C as recorded at OCV with 2 M methanol) and a suitable area specific resistance (0.15 Ohm cm2 for s-PEEK vs. 0.22 Ohm cm2 for Nafion® 115). PMID:26610582

Platinum supported on titanium–ruthenium oxide is a remarkably stable electrocatayst for hydrogen fuel cell vehicles

PubMed Central

Parrondo, Javier; Han, Taehee; Niangar, Ellazar; Wang, Chunmei; Dale, Nilesh; Adjemian, Kev; Ramani, Vijay

2014-01-01

We report a unique and highly stable electrocatalyst—platinum (Pt) supported on titanium–ruthenium oxide (TRO)—for hydrogen fuel cell vehicles. The Pt/TRO electrocatalyst was exposed to stringent accelerated test protocols designed to induce degradation and failure mechanisms identical to those seen during extended normal operation of a fuel cell automobile—namely, support corrosion during vehicle startup and shutdown, and platinum dissolution during vehicle acceleration and deceleration. These experiments were performed both ex situ (on supports and catalysts deposited onto a glassy carbon rotating disk electrode) and in situ (in a membrane electrode assembly). The Pt/TRO was compared against a state-of-the-art benchmark catalyst—Pt supported on high surface-area carbon (Pt/HSAC). In ex situ tests, Pt/TRO lost only 18% of its initial oxygen reduction reaction mass activity and 3% of its oxygen reduction reaction-specific activity, whereas the corresponding losses for Pt/HSAC were 52% and 22%. In in situ-accelerated degradation tests performed on membrane electrode assemblies, the loss in cell voltage at 1 A · cm−2 at 100% RH was a negligible 15 mV for Pt/TRO, whereas the loss was too high to permit operation at 1 A · cm−2 for Pt/HSAC. We clearly show that electrocatalyst support corrosion induced during fuel cell startup and shutdown is a far more potent failure mode than platinum dissolution during fuel cell operation. Hence, we posit that the need for a highly stable support (such as TRO) is paramount. Finally, we demonstrate that the corrosion of carbon present in the gas diffusion layer of the fuel cell is only of minor concern. PMID:24367118

Platinum supported on titanium-ruthenium oxide is a remarkably stable electrocatayst for hydrogen fuel cell vehicles.

PubMed

Parrondo, Javier; Han, Taehee; Niangar, Ellazar; Wang, Chunmei; Dale, Nilesh; Adjemian, Kev; Ramani, Vijay

2014-01-07

We report a unique and highly stable electrocatalyst-platinum (Pt) supported on titanium-ruthenium oxide (TRO)-for hydrogen fuel cell vehicles. The Pt/TRO electrocatalyst was exposed to stringent accelerated test protocols designed to induce degradation and failure mechanisms identical to those seen during extended normal operation of a fuel cell automobile-namely, support corrosion during vehicle startup and shutdown, and platinum dissolution during vehicle acceleration and deceleration. These experiments were performed both ex situ (on supports and catalysts deposited onto a glassy carbon rotating disk electrode) and in situ (in a membrane electrode assembly). The Pt/TRO was compared against a state-of-the-art benchmark catalyst-Pt supported on high surface-area carbon (Pt/HSAC). In ex situ tests, Pt/TRO lost only 18% of its initial oxygen reduction reaction mass activity and 3% of its oxygen reduction reaction-specific activity, whereas the corresponding losses for Pt/HSAC were 52% and 22%. In in situ-accelerated degradation tests performed on membrane electrode assemblies, the loss in cell voltage at 1 A · cm(-2) at 100% RH was a negligible 15 mV for Pt/TRO, whereas the loss was too high to permit operation at 1 A · cm(-2) for Pt/HSAC. We clearly show that electrocatalyst support corrosion induced during fuel cell startup and shutdown is a far more potent failure mode than platinum dissolution during fuel cell operation. Hence, we posit that the need for a highly stable support (such as TRO) is paramount. Finally, we demonstrate that the corrosion of carbon present in the gas diffusion layer of the fuel cell is only of minor concern.

Validation of the analytical methods in the LWR code BOXER for gadolinium-loaded fuel pins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paratte, J.M.; Arkuszewski, J.J.; Kamboj, B.K.

1990-01-01

Due to the very high absorption occurring in gadolinium-loaded fuel pins, calculations of lattices with such pins present are a demanding test of the analysis methods in light water reactor (LWR) cell and assembly codes. Considerable effort has, therefore, been devoted to the validation of code methods for gadolinia fuel. The goal of the work reported in this paper is to check the analysis methods in the LWR cell/assembly code BOXER and its associated cross-section processing code ETOBOX, by comparison of BOXER results with those from a very accurate Monte Carlo calculation for a gadolinium benchmark problem. Initial results ofmore » such a comparison have been previously reported. However, the Monte Carlo calculations, done with the MCNP code, were performed at Los Alamos National Laboratory using ENDF/B-V data, while the BOXER calculations were performed at the Paul Scherrer Institute using JEF-1 nuclear data. This difference in the basic nuclear data used for the two calculations, caused by the restricted nature of these evaluated data files, led to associated uncertainties in a comparison of the results for methods validation. In the joint investigations at the Georgia Institute of Technology and PSI, such uncertainty in this comparison was eliminated by using ENDF/B-V data for BOXER calculations at Georgia Tech.« less

Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

NASA Technical Reports Server (NTRS)

Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

1972-01-01

A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The range of the previous experimental investigations has been expanded to include the reactivity effects of:(1) surrounding the reactor with 15.24 cm (6 in.) of polyethylene, (2) reducing the heights of a portion of the upper and lower axial reflectors by factors of 2 and 4, (3) adding 45 kg of W to the core uniformly in two steps, (4) adding 9.54 kg of Ta to the core uniformly, and (5) inserting 2.3 kg of polyethylene into the core proper and determining the effect of a Ta addition on the polyethylene worth.

JEFF-3.1, ENDF/B-VII and JENDL-3.3 Critical Assemblies Benchmarking With the Monte Carlo Code TRIPOLI

NASA Astrophysics Data System (ADS)

Sublet, Jean-Christophe

2008-02-01

ENDF/B-VII.0, the first release of the ENDF/B-VII nuclear data library, was formally released in December 2006. Prior to this event the European JEFF-3.1 nuclear data library was distributed in April 2005, while the Japanese JENDL-3.3 library has been available since 2002. The recent releases of these neutron transport libraries and special purpose files, the updates of the processing tools and the significant progress in computer power and potency, allow today far better leaner Monte Carlo code and pointwise library integration leading to enhanced benchmarking studies. A TRIPOLI-4.4 critical assembly suite has been set up as a collection of 86 benchmarks taken principally from the International Handbook of Evaluated Criticality Benchmarks Experiments (2006 Edition). It contains cases for a variety of U and Pu fuels and systems, ranging from fast to deep thermal solutions and assemblies. It covers cases with a variety of moderators, reflectors, absorbers, spectra and geometries. The results presented show that while the most recent library ENDF/B-VII.0, which benefited from the timely development of JENDL-3.3 and JEFF-3.1, produces better overall results, it suggest clearly also that improvements are still needed. This is true in particular in Light Water Reactor applications for thermal and epithermal plutonium data for all libraries and fast uranium data for JEFF-3.1 and JENDL-3.3. It is also true to state that other domains, in which Monte Carlo code are been used, such as astrophysics, fusion, high-energy or medical, radiation transport in general benefit notably from such enhanced libraries. It is particularly noticeable in term of the number of isotopes, materials available, the overall quality of the data and the much broader energy range for which evaluated (as opposed to modeled) data are available, spanning from meV to hundreds of MeV. In pointing out the impact of the different nuclear data at the library but also the isotopic levels one could not help noticing the importance and difference of the compensating effects that result from their single usage. Library differences are still important but tend to diminish due to the ever increasing and beneficial worldwide collaboration in the field of nuclear data measurement and evaluations.

Compact propane fuel processor for auxiliary power unit application

NASA Astrophysics Data System (ADS)

Dokupil, M.; Spitta, C.; Mathiak, J.; Beckhaus, P.; Heinzel, A.

With focus on mobile applications a fuel cell auxiliary power unit (APU) using liquefied petroleum gas (LPG) is currently being developed at the Centre for Fuel Cell Technology (Zentrum für BrennstoffzellenTechnik, ZBT gGmbH). The system is consisting of an integrated compact and lightweight fuel processor and a low temperature PEM fuel cell for an electric power output of 300 W. This article is presenting the current status of development of the fuel processor which is designed for a nominal hydrogen output of 1 k Wth,H2 within a load range from 50 to 120%. A modular setup was chosen defining a reformer/burner module and a CO-purification module. Based on the performance specifications, thermodynamic simulations, benchmarking and selection of catalysts the modules have been developed and characterised simultaneously and then assembled to the complete fuel processor. Automated operation results in a cold startup time of about 25 min for nominal load and carbon monoxide output concentrations below 50 ppm for steady state and dynamic operation. Also fast transient response of the fuel processor at load changes with low fluctuations of the reformate gas composition have been achieved. Beside the development of the main reactors the transfer of the fuel processor to an autonomous system is of major concern. Hence, concepts for packaging have been developed resulting in a volume of 7 l and a weight of 3 kg. Further a selection of peripheral components has been tested and evaluated regarding to the substitution of the laboratory equipment.

S/sub n/ analysis of the TRX metal lattices with ENDF/B version III data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeler, F.J.; Pearlstein, S.

1975-03-01

Two critical assemblies, designated as thermal-reactor benchmarks TRX-1 and TRX-2 for ENDF/B data testing, were analyzed using the one-dimensional S/sub n/-theory code SCAMP. The two assemblies were simple lattices of aluminum-clad, uranium-metal fuel rods in triangular arrays with D$sub 2$O as moderator and reflector. The fuel was low-enriched (1.3 percent $sup 235$U), 0.387-inch in diameter and had an active height of 48 inches. The volume ratio of water to uranium was 2.35 for the TRX-1 lattice and 4.02 for TRX-2. Full-core S/sub n/ calculations based on Version III data were performed for these assemblies and the results obtained were comparedmore » with the measured values of the multiplication factors, the ratio of epithermal-to-thermal neutron capture in $sup 238$U, the ratio of epithermal-to-thermal fission in $sup 235$U, the ratio of $sup 238$U fission to $sup 235$U fission, and the ratio of capture in $sup 238$U to fission in $sup 235$U. Reaction rates were obtained from a central region of the full- core problems. Multigroup cross sections for the reactor calculation were obtained from S/sub n/ cell calculations with resonance self-shielding calculated using the RABBLE treatment. The results of the analyses are generally consistent with results obtained by other investigators. (auth)« less

Fuel rod assembly to manifold attachment

DOEpatents

Donck, Harry A.; Veca, Anthony R.; Snyder, Jr., Harold J.

1980-01-01

A fuel element is formed with a plurality of fuel rod assemblies detachably connected to an overhead support with each of the fuel rod assemblies having a gas tight seal with the support to allow internal fission gaseous products to flow without leakage from the fuel rod assemblies into a vent manifold passageway system on the support. The upper ends of the fuel rod assemblies are located at vertically extending openings in the support and upper threaded members are threaded to the fuel rod assemblies to connect the latter to the support. The preferred threaded members are cap nuts having a dome wall encircling an upper threaded end on the fuel rod assembly and having an upper sealing surface for sealing contact with the support. Another and lower seal is achieved by abutting a sealing surface on each fuel rod assembly with the support. A deformable portion on the cap nut locks the latter against inadvertent turning off the fuel rod assembly. Orienting means on the fuel rod and support primarily locates the fuel rods azimuthally for reception of a deforming tool for the cap nut. A cross port in the fuel rod end plug discharges into a sealed annulus within the support, which serves as a circumferential chamber, connecting the manifold gas passageways in the support.

Fuel injection assembly for use in turbine engines and method of assembling same

DOEpatents

Berry, Jonathan Dwight; Johnson, Thomas Edward; York, William David; Uhm, Jong Ho

2015-12-15

A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes an end cover, an endcap assembly, a fluid supply chamber, and a plurality of tube assemblies positioned at the endcap assembly. Each of the tube assemblies includes housing having a fuel plenum and a cooling fluid plenum. The cooling fluid plenum is positioned downstream from the fuel plenum and separated from the fuel plenum by an intermediate wall. The plurality of tube assemblies also include a plurality of tubes that extends through the housing. Each of the plurality of tubes is coupled in flow communication with the fluid supply chamber and a combustion chamber positioned downstream from the tube assembly. The plurality of tube assemblies further includes an aft plate at a downstream end of the cooling fluid plenum. The plate includes at least one aperture.

Control assembly for controlling a fuel cell system during shutdown and restart

DOEpatents

Venkataraman, Ramki; Berntsen, George; Carlson, Glenn L.; Farooque, Mohammad; Beachy, Dan; Peterhans, Stefan; Bischoff, Manfred

2010-06-15

A fuel cell system and method in which the fuel cell system receives and an input oxidant gas and an input fuel gas, and in which a fuel processing assembly is provided and is adapted to at least humidify the input fuel gas which is to be supplied to the anode of the fuel cell of the system whose cathode receives the oxidant input gas via an anode oxidizing assembly which is adapted to couple the output of the anode of the fuel cell to the inlet of the cathode of the fuel cell during normal operation, shutdown and restart of the fuel cell system, and in which a control assembly is further provided and is adapted to respond to shutdown of the fuel cell system during which input fuel gas and input oxidant gas cease to be received by the fuel cell system, the control assembly being further adapted to, when the fuel cell system is shut down: control the fuel cell system so as to enable a purging gas to be able to flow through the fuel processing assembly to remove humidified fuel gas from the processing assembly and to enable a purging gas to be able to flow through the anode of the fuel cell.

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lell, R. M.; Morman, J. A.; Schaefer, R.W.

ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide,more » U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE PAGES

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.; ...

2017-08-02

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

Fuel injection assembly for gas turbine engine combustor

NASA Technical Reports Server (NTRS)

Candy, Anthony J. (Inventor); Glynn, Christopher C. (Inventor); Barrett, John E. (Inventor)

2002-01-01

A fuel injection assembly for a gas turbine engine combustor, including at least one fuel stem, a plurality of concentrically disposed tubes positioned within each fuel stem, wherein a cooling supply flow passage, a cooling return flow passage, and a tip fuel flow passage are defined thereby, and at least one fuel tip assembly connected to each fuel stem so as to be in flow communication with the flow passages, wherein an active cooling circuit for each fuel stem and fuel tip assembly is maintained by providing all active fuel through the cooling supply flow passage and the cooling return flow passage during each stage of combustor operation. The fuel flowing through the active cooling circuit is then collected so that a predetermined portion thereof is provided to the tip fuel flow passage for injection by the fuel tip assembly.

76 FR 12825 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Confirmation of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-09

... definitions for Damaged Fuel Assembly and Transfer Operations; add definitions for Fuel Class and Reconstituted Fuel Assembly; add Combustion Engineering 16x16 class fuel assemblies as authorized contents...

Locking support for nuclear fuel assemblies

DOEpatents

Ledin, Eric

1980-01-01

A locking device for supporting and locking a nuclear fuel assembly within a cylindrical bore formed by a support plate, the locking device including a support and locking sleeve having upwardly extending fingers forming wedge shaped contact portions arranged for interaction between an annular tapered surface on the fuel assembly and the support plate bore as well as downwardly extending fingers having wedge shaped contact portions arranged for interaction between an annularly tapered surface on the support plate bore and the fuel assembly whereby the sleeve tends to support and lock the fuel assembly in place within the bore by its own weight while facilitating removal and/or replacement of the fuel assembly.

77 FR 70114 - Airworthiness Directives; Cessna Aircraft Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-23

... assemblies, which were caused by the fuel return line assembly rubbing against the right steering tube assembly during full rudder pedal actuation. This AD requires you to inspect the fuel return line assembly... the fuel return line assembly and both the right steering tube assembly and the airplane structure...

QUAD+ BWR Fuel Assembly demonstration program at Browns Ferry plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doshi, P.K.; Mayhue, L.T.; Robert, J.T.

1984-04-01

The QUAD+ fuel assembly is an improved BWR fuel assembly designed and manufactured by Westinghouse Electric Corporation. The design features a water cross separating four fuel minibundles in an integral channel. A demonstration program for this fuel design is planned for late 1984 in cycle 6 of Browns Ferry 2, a TVA plant. Objectives for the design of the QUAD+ demonstration assemblies are compatibility in performance and transparency in safety analysis with the feed fuel. These objectives are met. Inspections of the QUAD+ demonstration assemblies are planned at each refueling outage.

Development and Experimental Benchmark of Simulations to Predict Used Nuclear Fuel Cladding Temperatures during Drying and Transfer Operations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greiner, Miles

Radial hydride formation in high-burnup used fuel cladding has the potential to radically reduce its ductility and suitability for long-term storage and eventual transport. To avoid this formation, the maximum post-reactor temperature must remain sufficiently low to limit the cladding hoop stress, and so that hydrogen from the existing circumferential hydrides will not dissolve and become available to re-precipitate into radial hydrides under the slow cooling conditions during drying, transfer and early dry-cask storage. The objective of this research is to develop and experimentallybenchmark computational fluid dynamics simulations of heat transfer in post-pool-storage drying operations, when high-burnup fuel cladding ismore » likely to experience its highest temperature. These benchmarked tools can play a key role in evaluating dry cask storage systems for extended storage of high-burnup fuels and post-storage transportation, including fuel retrievability. The benchmarked tools will be used to aid the design of efficient drying processes, as well as estimate variations of surface temperatures as a means of inferring helium integrity inside the canister or cask. This work will be conducted effectively because the principal investigator has experience developing these types of simulations, and has constructed a test facility that can be used to benchmark them.« less

Method for shearing spent nuclear fuel assemblies

DOEpatents

Weil, Bradley S.; Watson, Clyde D.

1977-01-01

A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

Feebates and Fuel Economy Standards: Impacts on Fuel Use in Light-Duty Vehicles and Greenhouse Gas Emissions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greene, David L

2011-01-01

This study evaluates the potential impacts of a national feebate system, a market-based policy that consists of graduated fees on low-fuel-economy (or high-emitting) vehicles and rebates for high-fuel-economy (or lowemitting) vehicles. In their simplest form, feebate systems operate under three conditions: a benchmark divides all vehicles into two categories-those charged fees and those eligible for rebates; the sizes of the fees and rebates are a function of a vehicle's deviation from its benchmark; and placement of the benchmark ensures revenue neutrality or a desired level of subsidy or revenue. A model developed by the University of California for the Californiamore » Air Resources Board was revised and used to estimate the effects of six feebate structures on fuel economy and sales of new light-duty vehicles, given existing and anticipated future fuel economy and emission standards. These estimates for new vehicles were then entered into a vehicle stock model that simulated the evolution of the entire vehicle stock. The results indicate that feebates could produce large, additional reductions in emissions and fuel consumption, in large part by encouraging market acceptance of technologies with advanced fuel economy, such as hybrid electric vehicles.« less

Fuel cell sub-assembly

DOEpatents

Chi, Chang V.

1983-01-01

A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

DOEpatents

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

Improved nuclear fuel assembly grid spacer

DOEpatents

Marshall, John; Kaplan, Samuel

1977-01-01

An improved fuel assembly grid spacer and method of retaining the basic fuel rod support elements in position within the fuel assembly containment channel. The improvement involves attachment of the grids to the hexagonal channel and of forming the basic fuel rod support element into a grid structure, which provides a design which is insensitive to potential channel distortion (ballooning) at high fluence levels. In addition the improved method eliminates problems associated with component fabrication and assembly.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Since much of the fuel is very old, Cerenkov viewing devices are often not very useful even for the top layer. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times canmore » exist in any given pool. A spent fuel neutron counting tool consisting of a fission chamber, SFNC, has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups to levels up 11,000 MWd/t, the existing signal processing software of the tool was found to fail due to non-linearity of the source term with burnup. A new Graphical User Interface software package based on the LabVIEW platform was developed to predict expected neutron signals covering all ranges of burnups and cooling times and establish maps of expected signals at various pool locations. The algorithm employed in the software uses a set of transfer functions in a 47-energy group structure which are coupled with a 47-energy group neutron source spectrum based on various cooling times and burnups for each of the two enrichment levels. The database of the software consists of these transfer functions for the three different inter-assembly pitches that the fuel is stored in at the site. The transfer functions were developed for a 6 by 6 matrix of fuel assemblies with the detector placed at the center surrounded by four near neighbors, eight next nearest neighbors and so on for the 36 assemblies. These calculations were performed using Monte Carlo radiation transport methods. The basic methodology consisted of starting sources in each of the assemblies and tallying the contribution to the detector by a single neutron in each of the 47 energy groups used. Thus for the single existing symmetric pitch in the pools, where the vertical and horizontal separations are equal, only 6 sets of transfer functions are required. For the two asymmetrical pitches, nine sets of transfer functions are stored. In addition, source spectra at burnups ranging from 4000 to 20000 MWd/t and cooling times up to 40 years are stored. These source terms were established based on CANDU 37-rod fuel that is very similar to the Atucha fuel. Linear interpolation is used by the software for both burnup and cooling time to establish source terms at any intermediate condition. Using the burnup, cooling time and initial enrichment of the surrounding assemblies a set of source strengths in the 47-group structure for each of the 36 assemblies is established and multiplied group-wise with the appropriate transfer function set. The grand total over the 47 groups for all 36 assemblies is the predicted signal at the detector. The software was initially calibrated against a set of typically 5-6 measurements chosen from among the measured data at each level of the six pools and calibration factors were established. The set used for calibration is chosen such that it is fairly representative of the range of spent fuel assembly characteristics present in each level. Once established, these calibration factors can be repeatedly used for verification purposes. Recalibration will be required if the hardware or pool configurations has changed. It will also be required if a long enough time has elapsed since they were established thus making a cooling time correction necessary. The objective of the inspection is to detect missing fuel from one or more nearest neighbors of the detector. During the verification mode of the software, the predicted and measured signals are compared and the inspector is alerted if the difference between the two signals is beyond a set tolerance limit. Based on the uncertainties associated with both the calculations and measurements, a lower limit of the tolerance will be 15% with an upper limit of 20%. For the most part a 20% tolerance limit will be able to detect a missing assembly since in the vast majority of cases the drop in signal due to a single missing nearest neighbor assembly will be in the range 24-27%. The software was benchmarked against an extensive set of measured data taken at the site in 2004. Overall, 326 data points were examined and the prediction of the calibrated software was compared to the measurements within a set tolerance of ±20%. Of these, 283 of the predicted signals representing 87% of the total matched the measured data within ±10%. A further 27 or 8% were in the range of ±10-15% and 8 or 2.5% were in the range of ±15-20%. Thus, 97.5% of the data matched the measurements within the set tolerance limit of 20%, with 95% matching measured data with the lowest allowed tolerance limit of ±15%. The remaining 2.5% had measured signals that were very different from those at locations with very similar surrounding assemblies and the cause of these discrepancies could not be ascertained from the measurement logs. In summary, 97.5% of the predictions matched the measurements within the set 20% tolerance limit providing proof of the robustness of the software. This software package linked to SFNC will be deployed at the site and will enhance the capability of gross defect verification for the whole range of burnup, cooling time and initial enrichments of the spent fuel being discharged into the various pools at the Atucha-I reactor site.« less

Application of X-Ray Computer Tomography for Observing the Central Void Formations and the Fuel Pin Deformations of Irradiated FBR Fuel Assemblies

NASA Astrophysics Data System (ADS)

Katsuyama, Kozo; Nagamine, Tsuyoshi; Furuya, Hirotaka

2010-10-01

In order to observe the structural change in the interior of irradiated fuel assemblies, a non-destructive post-irradiation examination (PIE) technique using X-ray computer tomography (X-ray CT) was developed. This X-ray CT technique was applied to observe the central void formations and fuel pin deformations of fuel assemblies which had been irradiated at high linear heat rating. The central void sizes in all fuel pins were measured on five cross sections of the core fuel column as a parameter for evaluating fuel thermal performance. In addition, the fuel pin deformations were analyzed from X-ray CT images obtained along the axial direction of a fuel assembly at the same separation interval. A dependence of void size on the linear heat rating was seen in the fuel assembly irradiated at high linear heat rating. In addition, significant undulations of the fuel pin were observed along the axial direction, coinciding with the wrapping wire pitch in the core fuel column. Application of the developed technique should provide enhanced resolution of measurements and simplify fuel PIEs.

The Conceptual Design for a Fuel Assembly of a New Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryu, J-S.; Cho, Y-G.; Yoon, D-B.

2004-10-06

A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibrationmore » characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.« less

System and method for controlling a combustor assembly

DOEpatents

York, William David; Ziminsky, Willy Steve; Johnson, Thomas Edward; Stevenson, Christian Xavier

2013-03-05

A system and method for controlling a combustor assembly are disclosed. The system includes a combustor assembly. The combustor assembly includes a combustor and a fuel nozzle assembly. The combustor includes a casing. The fuel nozzle assembly is positioned at least partially within the casing and includes a fuel nozzle. The fuel nozzle assembly further defines a head end. The system further includes a viewing device configured for capturing an image of at least a portion of the head end, and a processor communicatively coupled to the viewing device, the processor configured to compare the image to a standard image for the head end.

Swelling-resistant nuclear fuel

DOEpatents

Arsenlis, Athanasios [Hayward, CA; Satcher, Jr., Joe; Kucheyev, Sergei O [Oakland, CA

2011-12-27

A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

40 CFR 1060.102 - What permeation emission control requirements apply for fuel lines?

Code of Federal Regulations, 2011 CFR

2011-07-01

... assemblies as aggregated systems that include multiple sections of fuel line with connectors and fittings. For example, you may certify fuel lines for portable marine fuel tanks as assemblies of fuel hose, primer bulbs, and self-sealing end connections. The length of such an assembly must not be longer than a...

40 CFR 1060.102 - What permeation emission control requirements apply for fuel lines?

Code of Federal Regulations, 2010 CFR

2010-07-01

... assemblies as aggregated systems that include multiple sections of fuel line with connectors and fittings. For example, you may certify fuel lines for portable marine fuel tanks as assemblies of fuel hose, primer bulbs, and self-sealing end connections. The length of such an assembly must not be longer than a...

Fabrication of fuel pin assemblies, phase 3

NASA Technical Reports Server (NTRS)

Keeton, A. R.; Stemann, L. G.

1972-01-01

Five full size and eight reduced length fuel pins were fabricated for irradiation testing to evaluate design concepts for a fast spectrum lithium cooled compact space power reactor. These assemblies consisted of uranium mononitride fuel pellets encased in a T-111 (Ta-8W-2Hf) clad with a tungsten barrier separating fuel and clad. Fabrication procedures were fully qualified by process development and assembly qualification tests. Detailed specifications and procedures were written for the fabrication and assembly of prototype fuel pins.

Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pond, R.B.; Matos, J.E.

1996-05-01

As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, aremore » not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.« less

LMFBR fuel assembly design for HCDA fuel dispersal

DOEpatents

Lacko, Robert E.; Tilbrook, Roger W.

1984-01-01

A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

Methods of conditioning direct methanol fuel cells

DOEpatents

Rice, Cynthia; Ren, Xiaoming; Gottesfeld, Shimshon

2005-11-08

Methods for conditioning the membrane electrode assembly of a direct methanol fuel cell ("DMFC") are disclosed. In a first method, an electrical current of polarity opposite to that used in a functioning direct methanol fuel cell is passed through the anode surface of the membrane electrode assembly. In a second method, methanol is supplied to an anode surface of the membrane electrode assembly, allowed to cross over the polymer electrolyte membrane of the membrane electrode assembly to a cathode surface of the membrane electrode assembly, and an electrical current of polarity opposite to that in a functioning direct methanol fuel cell is drawn through the membrane electrode assembly, wherein methanol is oxidized at the cathode surface of the membrane electrode assembly while the catalyst on the anode surface is reduced. Surface oxides on the direct methanol fuel cell anode catalyst of the membrane electrode assembly are thereby reduced.

Flashback resistant pre-mixer assembly

DOEpatents

Laster, Walter R [Oviedo, FL; Gambacorta, Domenico [Oviedo, FL

2012-02-14

A pre-mixer assembly associated with a fuel supply system for mixing of air and fuel upstream from a main combustion zone in a gas turbine engine. The pre-mixer assembly includes a swirler assembly disposed about a fuel injector of the fuel supply system and a pre-mixer transition member. The swirler assembly includes a forward end defining an air inlet and an opposed aft end. The pre-mixer transition member has a forward end affixed to the aft end of the swirler assembly and an opposed aft end defining an outlet of the pre-mixer assembly. The aft end of the pre-mixer transition member is spaced from a base plate such that a gap is formed between the aft end of the pre-mixer transition member and the base plate for permitting a flow of purge air therethrough to increase a velocity of the air/fuel mixture exiting the pre-mixer assembly.

Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

DOEpatents

Bradley, John G.

1982-01-01

A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

NEAMS Update. Quarterly Report for October - December 2011.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bradley, K.

2012-02-16

The Advanced Modeling and Simulation Office within the DOE Office of Nuclear Energy (NE) has been charged with revolutionizing the design tools used to build nuclear power plants during the next 10 years. To accomplish this, the DOE has brought together the national laboratories, U.S. universities, and the nuclear energy industry to establish the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program. The mission of NEAMS is to modernize computer modeling of nuclear energy systems and improve the fidelity and validity of modeling results using contemporary software environments and high-performance computers. NEAMS will create a set of engineering-level codes aimedmore » at designing and analyzing the performance and safety of nuclear power plants and reactor fuels. The truly predictive nature of these codes will be achieved by modeling the governing phenomena at the spatial and temporal scales that dominate the behavior. These codes will be executed within a simulation environment that orchestrates code integration with respect to spatial meshing, computational resources, and execution to give the user a common 'look and feel' for setting up problems and displaying results. NEAMS is building upon a suite of existing simulation tools, including those developed by the federal Scientific Discovery through Advanced Computing and Advanced Simulation and Computing programs. NEAMS also draws upon existing simulation tools for materials and nuclear systems, although many of these are limited in terms of scale, applicability, and portability (their ability to be integrated into contemporary software and hardware architectures). NEAMS investments have directly and indirectly supported additional NE research and development programs, including those devoted to waste repositories, safeguarded separations systems, and long-term storage of used nuclear fuel. NEAMS is organized into two broad efforts, each comprising four elements. The quarterly highlights October-December 2011 are: (1) Version 1.0 of AMP, the fuel assembly performance code, was tested on the JAGUAR supercomputer and released on November 1, 2011, a detailed discussion of this new simulation tool is given; (2) A coolant sub-channel model and a preliminary UO{sub 2} smeared-cracking model were implemented in BISON, the single-pin fuel code, more information on how these models were developed and benchmarked is given; (3) The Object Kinetic Monte Carlo model was implemented to account for nucleation events in meso-scale simulations and a discussion of the significance of this advance is given; (4) The SHARP neutronics module, PROTEUS, was expanded to be applicable to all types of reactors, and a discussion of the importance of PROTEUS is given; (5) A plan has been finalized for integrating the high-fidelity, three-dimensional reactor code SHARP with both the systems-level code RELAP7 and the fuel assembly code AMP. This is a new initiative; (6) Work began to evaluate the applicability of AMP to the problem of dry storage of used fuel and to define a relevant problem to test the applicability; (7) A code to obtain phonon spectra from the force-constant matrix for a crystalline lattice has been completed. This important bridge between subcontinuum and continuum phenomena is discussed; (8) Benchmarking was begun on the meso-scale, finite-element fuels code MARMOT to validate its new variable splitting algorithm; (9) A very computationally demanding simulation of diffusion-driven nucleation of new microstructural features has been completed. An explanation of the difficulty of this simulation is given; (10) Experiments were conducted with deformed steel to validate a crystal plasticity finite-element code for bodycentered cubic iron; (11) The Capability Transfer Roadmap was completed and published as an internal laboratory technical report; (12) The AMP fuel assembly code input generator was integrated into the NEAMS Integrated Computational Environment (NiCE). More details on the planned NEAMS computing environment is given; and (13) The NEAMS program website (neams.energy.gov) is nearly ready to launch.« less

GEN-IV Benchmarking of Triso Fuel Performance Models under accident conditions modeling input data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise Paul

This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: • The modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release. • The modeling of the AGR-1 and HFR-EU1bis safety testing experiments. •more » The comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from “Case 5” of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. “Case 5” of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to “effects of the numerical calculation method rather than the physical model” [IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary. 09/2016: Tables 6 and 8 updated. AGR-2 input data added« less

Generation IV benchmarking of TRISO fuel performance models under accident conditions: Modeling input data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.

2014-09-01

This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: the modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release; the modeling of the AGR-1 and HFR-EU1bis safety testing experiments; and, the comparisonmore » of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from ''Case 5'' of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. ''Case 5'' of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to ''effects of the numerical calculation method rather than the physical model''[IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary.« less

Interface ring for gas turbine fuel nozzle assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, Timothy A.; Schilp, Reinhard

A gas turbine combustor assembly including a combustor liner and a plurality of fuel nozzle assemblies arranged in an annular array extending within the combustor liner. The fuel nozzle assemblies each include fuel nozzle body integral with a swirler assembly, and the swirler assemblies each include a bellmouth structure to turn air radially inwardly for passage into the swirler assemblies. A radially outer removed portion of each of the bellmouth structures defines a periphery diameter spaced from an inner surface of the combustor liner, and an interface ring is provided extending between the combustor liner and the removed portions ofmore » the bellmouth structures at the periphery diameter.« less

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE PAGES

Ham, Y.; Kerr, P.; Sitaraman, S.; ...

2016-05-05

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Y.S.; Kerr, P.; Sitaraman, S.

The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported themore » successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)« less

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Y.; Kerr, P.; Sitaraman, S.

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Inactive end cell assembly for fuel cells for improved electrolyte management and electrical contact

DOEpatents

Yuh, Chao-Yi [New Milford, CT; Farooque, Mohammad [Danbury, CT; Johnsen, Richard [New Fairfield, CT

2007-04-10

An assembly for storing electrolyte in a carbonate fuel cell is provided. The combination of a soft, compliant and resilient cathode current collector and an inactive anode part including a foam anode in each assembly mitigates electrical contact loss during operation of the fuel cell stack. In addition, an electrode reservoir in the positive end assembly and an electrode sink in the negative end assembly are provided, by which ribbed and flat cathode members inhibit electrolyte migration in the fuel cell stack.

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE PAGES

Jiang, Hao; Wang, Jy-An John

2016-10-14

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

System and method for injecting fuel

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward

2012-12-04

According to various embodiments, a system includes a staggered multi-nozzle assembly. The staggered multi-nozzle assembly includes a first fuel nozzle having a first axis and a first flow path extending to a first downstream end portion, wherein the first fuel nozzle has a first non-circular perimeter at the first downstream end portion. The staggered multi-nozzle assembly also includes a second fuel nozzle having a second axis and a second flow path extending to a second downstream end portion, wherein the first and second downstream end portions are axially offset from one another relative to the first and second axes. The staggered multi-nozzle assembly further includes a cap member disposed circumferentially about at least the first and second fuel nozzles to assemble the staggered multi-nozzle assembly.

Fuel assembly for nuclear reactors

DOEpatents

Creagan, Robert J.; Frisch, Erling

1977-01-01

A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

Spent fuel burnup estimation by Cerenkov glow intensity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki

1994-10-01

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to themore » declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.« less

Lead Slowing-Down Spectrometry for Spent Fuel Assay: FY11 Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Warren, Glen A.; Casella, Andrew M.; Haight, R. C.

2011-08-01

Executive Summary Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) from next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration to study the feasibility of Lead Slowing Down Spectroscopy (LSDS). This technique is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than themore » approximately 10% typical of today’s confirmatory assay methods. This document is a progress report for FY2011 collaboration activities. Progress made by the collaboration in FY2011 continues to indicate the promise of LSDS techniques applied to used fuel. PNNL developed an empirical model based on calibration of the LSDS to responses generated from well-characterized used fuel. The empirical model demonstrated the potential for the direct and independent assay of the sum of the masses of 239Pu and 241Pu to within approximately 3% over a wide used fuel parameter space. Similar results were obtained using a perturbation approach developed by LANL. Benchmark measurements have been successfully conducted at LANL and at RPI using their respective LSDS instruments. The ISU and UNLV collaborative effort is focused on the fabrication and testing of prototype fission chambers lined with ultra-depleted 238U and 232Th, and uranium deposition on a stainless steel disc using spiked U3O8 from room temperature ionic liquid was successful, with improving thickness obtained. In FY2012, the collaboration plans a broad array of activities. PNNL will focus on optimizing its empirical model and minimizing its reliance on calibration data, as well continuing efforts on developing an analytical model. Additional measurements are planned at LANL and RPI. LANL measurements will include a Pu sample, which is expected to provide more counts at longer slowing-down times to help identify discrepancies between experimental data and MCNPX simulations. RPI measurements will include the assay of an entire fresh fuel assembly for the study of self-shielding effects as well as the ability to detect diversion by detecting a missing fuel pin in the fuel assembly. The development of threshold neutron sensors will continue, and UNLV will calibrate existing ultra-depleted uranium deposits at ISU.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wiebe, David J; Fox, Timothy A

A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passagemore » is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.« less

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, M.L.; Rosenstein, R.G.

1998-10-13

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

Mox fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-05-15

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-07-17

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

1998-01-01

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

Update and evaluation of decay data for spent nuclear fuel analyses

NASA Astrophysics Data System (ADS)

Simeonov, Teodosi; Wemple, Charles

2017-09-01

Studsvik's approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL) and processed (ESTAR) sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources). Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

The reasons for the high power density of fuel cells fabricated with directly deposited membranes

NASA Astrophysics Data System (ADS)

Vierrath, Severin; Breitwieser, Matthias; Klingele, Matthias; Britton, Benjamin; Holdcroft, Steven; Zengerle, Roland; Thiele, Simon

2016-09-01

In a previous study, we reported that polymer electrolyte fuel cells prepared by direct membrane deposition (DMD) produced power densities in excess of 4 W/cm2. In this study, the underlying origins that give rise to these high power densities are investigated and reported. The membranes of high power, DMD-fabricated fuel cells are relatively thin (12 μm) compared to typical benchmark, commercially available membranes. Electrochemical impedance spectroscopy, at high current densities (2.2 A/cm2) reveals that mass transport resistance was half that of reference, catalyst-coated-membranes (CCM). This is attributed to an improved oxygen supply in the cathode catalyst layer by way of a reduced propensity of flooding, and which is facilitated by an enhancement in the back diffusion of water from cathode to anode through the thin directly deposited membrane. DMD-fabricated membrane-electrode-assemblies possess 50% reduction in ionic resistance (15 mΩcm2) compared to conventional CCMs, with contributions of 9 mΩcm2 for the membrane resistance and 6 mΩcm2 for the contact resistance of the membrane and catalyst layer ionomer. The improved mass transport is responsible for 90% of the increase in power density of the DMD fuel cell, while the reduced ionic resistance accounts for a 10% of the improvement.

Reduced size fuel cell for portable applications

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor); Clara, Filiberto (Inventor); Frank, Harvey A. (Inventor)

2004-01-01

A flat pack type fuel cell includes a plurality of membrane electrode assemblies. Each membrane electrode assembly is formed of an anode, an electrolyte, and an cathode with appropriate catalysts thereon. The anode is directly into contact with fuel via a wicking element. The fuel reservoir may extend along the same axis as the membrane electrode assemblies, so that fuel can be applied to each of the anodes. Each of the fuel cell elements is interconnected together to provide the voltage outputs in series.

Coolant mass flow equalizer for nuclear fuel

DOEpatents

Betten, Paul R.

1978-01-01

The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

Shock ignition of thermonuclear fuel with high areal density.

PubMed

Betti, R; Zhou, C D; Anderson, K S; Perkins, L J; Theobald, W; Solodov, A A

2007-04-13

A novel method by C. Zhou and R. Betti [Bull. Am. Phys. Soc. 50, 140 (2005)] to assemble and ignite thermonuclear fuel is presented. Massive cryogenic shells are first imploded by direct laser light with a low implosion velocity and on a low adiabat leading to fuel assemblies with large areal densities. The assembled fuel is ignited from a central hot spot heated by the collision of a spherically convergent ignitor shock and the return shock. The resulting fuel assembly features a hot-spot pressure greater than the surrounding dense fuel pressure. Such a nonisobaric assembly requires a lower energy threshold for ignition than the conventional isobaric one. The ignitor shock can be launched by a spike in the laser power or by particle beams. The thermonuclear gain can be significantly larger than in conventional isobaric ignition for equal driver energy.

Method of preparing gas tags for identification of single and multiple failures of nuclear reactor fuel assemblies

DOEpatents

McCormick, Norman J.

1976-01-01

For use in the identification of failed fuel assemblies in a nuclear reactor, the ratios of the tag gas isotopic concentrations are located on curved surfaces to enable the ratios corresponding to failure of a single fuel assembly to be distinguished from those formed from any combination of two or more failed assemblies.

Specifications for a coupled neutronics thermal-hydraulics SFR test case

NASA Astrophysics Data System (ADS)

Tassone, A.; Smirnov, A. D.; Tikhomirov, G. V.

2017-01-01

Coupling neutronics/thermal-hydraulics calculations for the design of nuclear reactors are a growing trend in the scientific community. This approach allows to properly represent the mutual feedbacks between the neutronic distribution and the thermal-hydraulics properties of the materials composing the reactor, details which are often lost when separate analysis are performed. In this work, a test case for a generation IV sodium-cooled fast reactor (SFR), based on the ASTRID concept developed by CEA, is proposed. Two sub-assemblies (SA) characterized by different fuel enrichment and layout are considered. Specifications for the test case are provided including geometrical data, material compositions, thermo-physical properties and coupling scheme details. Serpent and ANSYS-CFX are used as reference in the description of suitable inputs for the performing of the benchmark, but the use of other code combinations for the purpose of validation of the results is encouraged. The expected outcome of the test case are the axial distribution of volumetric power generation term (q‴), density and temperature for the fuel, the cladding and the coolant.

Turbine combustor with fuel nozzles having inner and outer fuel circuits

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward; Kim, Kwanwoo

2013-12-24

A combustor cap assembly for a turbine engine includes a combustor cap and a plurality of fuel nozzles mounted on the combustor cap. One or more of the fuel nozzles would include two separate fuel circuits which are individually controllable. The combustor cap assembly would be controlled so that individual fuel circuits of the fuel nozzles are operated or deliberately shut off to provide for physical separation between the flow of fuel delivered by adjacent fuel nozzles and/or so that adjacent fuel nozzles operate at different pressure differentials. Operating a combustor cap assembly in this fashion helps to reduce or eliminate the generation of undesirable and potentially harmful noise.

Fuel and oxidizer valve assembly employs single solenoid actuator

NASA Technical Reports Server (NTRS)

1966-01-01

Valve assembly simultaneously starts or stops the flow of oxidizer and fuel from separate inlet channels to reaction control motors. The assembly combines an oxidizer shutoff valve and a fuel shutoff valve which are mechanically linked and operated by a single high-speed solenoid actuator.

Modular fuel-cell stack assembly

DOEpatents

Patel, Pinakin

2010-07-13

A fuel cell assembly having a plurality of fuel cells arranged in a stack. An end plate assembly abuts the fuel cell at an end of said stack. The end plate assembly has an inlet area adapted to receive an exhaust gas from the stack, an outlet area and a passage connecting the inlet area and outlet area and adapted to carry the exhaust gas received at the inlet area from the inlet area to the outlet area. A further end plate assembly abuts the fuel cell at a further opposing end of the stack. The further end plate assembly has a further inlet area adapted to receive a further exhaust gas from the stack, a further outlet area and a further passage connecting the further inlet area and further outlet area and adapted to carry the further exhaust gas received at the further inlet area from the further inlet area to the further outlet area.

Method and apparatus for close packing of nuclear fuel assemblies

DOEpatents

Newman, Darrell F.

1993-01-01

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Method and apparatus for close packing of nuclear fuel assemblies

DOEpatents

Newman, D.F.

1993-03-30

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Pool-site fuel inspection and examination techniques applied by the Kraftwerk Union Aktiengesellschaft Fuel Service. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knaab, H.; Knecht, K.

The need for pool-site inspection and examination of fuel assemblies was recognized by Kraftwerk Union Aktiengesellschaft with the commissioning of the first nuclear power stations. A wet sipping method has demonstrated high reliability in detection of leaking fuel assemblies. The visual inspection system is a versatile tool. It can be supplemented by attaching devices for oxide thickness measurement or surface replication. Repair of leaking pressurized water reactor fuel assemblies has improved fuel utilization. Applied methods and typical results are described.

Piloted rich-catalytic lean-burn hybrid combustor

DOEpatents

Newburry, Donald Maurice

2002-01-01

A catalytic combustor assembly which includes, an air source, a fuel delivery means, a catalytic reactor assembly, a mixing chamber, and a means for igniting a fuel/air mixture. The catalytic reactor assembly is in fluid communication with the air source and fuel delivery means and has a fuel/air plenum which is coated with a catalytic material. The fuel/air plenum has cooling air conduits passing therethrough which have an upstream end. The upstream end of the cooling conduits is in fluid communication with the air source but not the fuel delivery means.

Some methods for achieving more efficient performance of fuel assemblies

NASA Astrophysics Data System (ADS)

Boltenko, E. A.

2014-07-01

More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.

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

NASA Astrophysics Data System (ADS)

Karahan, Aydın

2011-07-01

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other hand, in order to accommodate solid fission product swelling and to control fuel clad mechanical interaction of the stiffer fuel, the fuel smear density is reduced to 70%. In addition, plenum height is increased to accommodate for fission gases.

Assess How Changes in Fuel Cycle Operation Impact Safeguards

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen Joseph; Adigun, Babatunde John; Fugate, Michael Lynn

Since the beginning of commercial nuclear power generation in the 1960s, the ability of researchers to understand and control the isotopic content of spent fuel has improved. It is therefore not surprising that both fuel assembly design and fuel assembly irradiation optimization have improved over the past 50+ years. It is anticipated that the burnup and isotopics of the spent fuel should exhibit less variation over the decades as reactor operators irradiate each assembly to the optimum amount. In contrast, older spent fuel is anticipated to vary more in burnup and resulting isotopics for a given initial enrichment. Modern fuelmore » therefore should be more uniform in composition, and thus, measured safeguards results should be easier to interpret than results from older spent fuel. With spent fuel ponds filling up, interim and long-­term storage of spent fuel will need to be addressed. Additionally after long periods of storage, spent fuel is no longer self-­protecting and, as such, the IAEA will categorize it as more attractive; in approximately 20 years many of the assemblies from early commercial cores will no longer be considered self-­protecting. This study will assess how more recent changes in the reactor operation could impact the interpretation of safeguards measurements. The status quo for spent fuel assay in the safeguards context is that the overwhelming majority of spent fuel assemblies are not measured in a quantitative way except for those assemblies about to be loaded into a difficult or impossible to access location (dry storage or, in the future, a repository). In other words, when the assembly is still accessible to a state actor, or an insider, when it is cooling in a pool, the inspectorate does not have a measurement database that could assist them in re-­verifying the integrity of that assembly. The spent fuel safeguards regime would be strengthened if spent fuel assemblies were measured from discharge to loading into a difficult or impossible to access location. The primary driver for suggesting this shift in approach is the change in robotic technology and information technology in general. It should be possible, with minimal impact to the facility, to measure each assembly every time that it is moved in the pool, with the first measurements being made at discharge. The following conclusions were reached: The total neutron count rate can be accurately predicted at any future moment in time based upon the measured count rate at discharge, provided the initial enrichment and burnup of the assembly is known at discharge. It is expected that the total neutron count rate measured at discharge will be indicative of the initial enrichment and burnup of that assembly. If the automated robot were to focus on measuring the assemblies in the rack without moving them, the time available would increase immensely.« less

Analysis of gamma ray dose for dried up pond storing low enriched UO2 fuel

NASA Astrophysics Data System (ADS)

Nauchi, Yasushi; Suzuki, Motomu

2017-09-01

Gamma ray dose is calculated for loss of coolant accident in spent fuel pond (SFP) storing irradiated fuels used in light water reactors. Influence of modelling of fuel assemblies, source distributions, and loading fraction of fuel assemblies in the fuel rack on the dose are investigated.

Transportation Fuels and the Hydrogen Economy

NASA Astrophysics Data System (ADS)

Gabbard, Alex

2004-11-01

An energy analysis of transportation fuels is performed for comparing automobiles and fuels currently in the marketplace as real world benchmarks projected as "hydrogen economy" requirements. Comparisons are made for ideal case average energy values at Standard Temperature and Pressure (STP) at 20°C, 1 atmosphere with no loses. "Real world" benchmarks currently in the marketplace illuminate the challenges to be met if an equivalent "hydrogen economy" is to become reality. The idea of a "hydrogen economy" is that, at some time in the future, world energy needs will be supplied in part or totally from hydrogen; in part as compared to the current "petroleum economy" that is the source of most of the world's transportation fuels and only a portion of total energy use, or hydrogen as the source of all energy consumption.

Fuel Cell Technology Status Analysis Project: Partnership Opportunities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fact sheet describing the National Renewable Energy Laboratory's (NREL's) Fuel Cell Technology Status Analysis Project. NREL is seeking fuel cell industry partners from the United States and abroad to participate in an objective and credible analysis of commercially available fuel cell products to benchmark the current state of the technology and support industry growth.

FUEL ASSEMBLY SHAKER TEST SIMULATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.

This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refinedmore » to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through direct comparison of model results to recorded test results. This does not offer validation for the fuel assembly model in all conceivable cases, such as high kinetic energy shock cases where the fuel assembly might lift off the basket floor to strike to basket ceiling. This type of nonlinear behavior was not witnessed in testing, so the model does not have test data to be validated against.a basis for validation in cases that substantially alter the fuel assembly response range. This leads to a gap in knowledge that is identified through this modeling study. The SNL shaker testing loaded a surrogate fuel assembly with a certain set of artificially-generated time histories. One thing all the shock cases had in common was an elimination of low frequency components, which reduces the rigid body dynamic response of the system. It is not known if the SNL test cases effectively bound all highway transportation scenarios, or if significantly greater rigid body motion than was tested is credible. This knowledge gap could be filled through modeling the vehicle dynamics of a used fuel conveyance, or by collecting acceleration time history data from an actual conveyance under highway conditions.« less

Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ellis, Ronald James

The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) duringmore » cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.« less

Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

NASA Astrophysics Data System (ADS)

Hausladen, Paul; Iyengar, Anagha; Fabris, Lorenzo; Yang, Jinan; Hu, Jianwei; Blackston, Matthew

2017-09-01

Oak Ridge National Laboratory is developing a new capability to perform passive fast neutron emission tomography of spent nuclear fuel assemblies for the purpose of verifying their integrity for international safeguards applications. Most of the world's plutonium is contained in spent nuclear fuel, so it is desirable to detect the diversion of irradiated fuel rods from an assembly prior to its transfer to ``difficult to access'' storage, such as a dry cask or permanent repository, where re-verification is practically impossible. Nuclear fuel assemblies typically consist of an array of fuel rods that, depending on exposure in the reactor and consequent ingrowth of 244Cm, are spontaneous sources of as many as 109 neutrons s-1. Neutron emission tomography uses collimation to isolate neutron activity along ``lines of response'' through the assembly and, by combining many collimated views through the object, mathematically extracts the neutron emission from each fuel rod. This technique, by combining the use of fast neutrons -which can penetrate the entire fuel assembly -and computed tomography, is capable of detecting vacancies or substitutions of individual fuel rods. This paper will report on the physics design and component testing of the imaging system. This material is based upon work supported by the U.S. Department of Energy, Office of Defense Nuclear Nonproliferation Research and Development within the National Nuclear Security Administration, under Contract Number DE-AC05-00OR22725.

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andress, D.; Joy, D.S.; McLeod, N.B.

The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elementsmore » as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.« less

Fuel injection assembly for use in turbine engines and method of assembling same

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward

2015-03-24

A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes a plurality of tube assemblies, wherein each of the tube assemblies includes an upstream portion and a downstream portion. Each tube assembly includes a plurality of tubes that extend from the upstream portion to the downstream portion or from the upstream portion through the downstream portion. At least one injection system is coupled to at least one tube assembly of the plurality of tube assemblies. The injection system includes a fluid supply member that extends from a fluid source to the downstream portion of the tube assembly. The fluid supply member includes a first end portion located in the downstream portion of the tube assembly, wherein the first end portion has at least one first opening for channeling fluid through the tube assembly to facilitate reducing a temperature therein.

Sputter-deposited fuel cell membranes and electrodes

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R. (Inventor); Jeffries-Nakamura, Barbara (Inventor); Chun, William (Inventor); Ruiz, Ron P. (Inventor); Valdez, Thomas I. (Inventor)

2001-01-01

A method for preparing a membrane for use in a fuel cell membrane electrode assembly includes the steps of providing an electrolyte membrane, and sputter-depositing a catalyst onto the electrolyte membrane. The sputter-deposited catalyst may be applied to multiple sides of the electrolyte membrane. A method for forming an electrode for use in a fuel cell membrane electrode assembly includes the steps of obtaining a catalyst, obtaining a backing, and sputter-depositing the catalyst onto the backing. The membranes and electrodes are useful for assembling fuel cells that include an anode electrode, a cathode electrode, a fuel supply, and an electrolyte membrane, wherein the electrolyte membrane includes a sputter-deposited catalyst, and the sputter-deposited catalyst is effective for sustaining a voltage across a membrane electrode assembly in the fuel cell.

Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen Joseph; Fugate, Michael Lynn; Trellue, Holly Renee

2016-10-28

A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly tomore » confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.O. Bader

The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are releasedmore » from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies.« less

Benchmarking Data for the Proposed Signature of Used Fuel Casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rauch, Eric Benton

2016-09-23

A set of benchmarking measurements to test facets of the proposed extended storage signature was conducted on May 17, 2016. The measurements were designed to test the overall concept of how the proposed signature can be used to identify a used fuel cask based only on the distribution of neutron sources within the cask. To simulate the distribution, 4 Cf-252 sources were chosen and arranged on a 3x3 grid in 3 different patterns and raw neutron totals counts were taken at 6 locations around the grid. This is a very simplified test of the typical geometry studied previously in simulationmore » with simulated used nuclear fuel.« less

Fabrication of capsule assemblies, phase 3

NASA Technical Reports Server (NTRS)

Keeton, A. R.; Stemann, L. G.

1973-01-01

Thirteen capsule assemblies were fabricated for evaluation of fuel pin design concepts for a fast spectrum lithium cooled compact space power reactor. These instrumented assemblies were designed for real time test of prototype fuel pins. Uranium mononitride fuel pins were encased in AISI 304L stainless steel capsules. Fabrication procedures were fully qualified by process development and assembly qualification tests. Instrumentation reliability was achieved utilizing specially processed and closely controlled thermocouple hot zone fabrication and by thermal screening tests. Overall capsule reliability was achieved with an all electron beam welded assembly.

Nuclear core and fuel assemblies

DOEpatents

Downs, Robert E.

1981-01-01

A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, W.E.; Trapp, T.J.

1983-06-10

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, William E.; Trapp, Turner J.

1985-01-01

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-11-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark reactor physics data to support validation of the reactor physics codes used to design commercial reactor fuel elements in an enrichment range above the current 5% enrichment cap. A first set of critical experiments in the 7uPCX has been completed. More experiments are planned in the 7uPCX series. The critical experiments at Sandia National Laboratories are currently funded by the US Department of Energy Nuclear Criticality Safety Program (NCSP). The NCSP has committed to maintain the critical experiment capability at Sandia and to support the development of a critical experiments training course at the facility. The training course is intended to provide hands-on experiment experience for the training of new and re-training of practicing Nuclear Criticality Safety Engineers. The current plans are for the development of the course to continue through the first part of fiscal year 2011 with the development culminating is the delivery of a prototype of the course in the latter part of the fiscal year. The course will be available in fiscal year 2012.« less

Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly †

PubMed Central

De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

2016-01-01

Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

Fuel cell system including a unit for electrical isolation of a fuel cell stack from a manifold assembly and method therefor

DOEpatents

Kelley; Dana A. , Farooque; Mohammad , Davis; Keith

2007-10-02

A fuel cell system with improved electrical isolation having a fuel cell stack with a positive potential end and a negative potential, a manifold for use in coupling gases to and from a face of the fuel cell stack, an electrical isolating assembly for electrically isolating the manifold from the stack, and a unit for adjusting an electrical potential of the manifold such as to impede the flow of electrolyte from the stack across the isolating assembly.

Separator assembly for use in spent nuclear fuel shipping cask

DOEpatents

Bucholz, James A.

1983-01-01

A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fact sheet describing the National Renewable Energy Laboratory's (NREL's) Fuel Cell Technology Status Analysis Project. NREL is seeking fuel cell industry partners from the United States and abroad to participate in an objective and credible analysis of commercially available fuel cell products to benchmark the current state of the technology and support industry growth.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Carlsen, Brett W.; Dixon, Brent W.

Dynamic fuel cycle simulation tools are intended to model holistic transient nuclear fuel cycle scenarios. As with all simulation tools, fuel cycle simulators require verification through unit tests, benchmark cases, and integral tests. Model validation is a vital aspect as well. Although compara-tive studies have been performed, there is no comprehensive unit test and benchmark library for fuel cycle simulator tools. The objective of this paper is to identify the must test functionalities of a fuel cycle simulator tool within the context of specific problems of interest to the Fuel Cycle Options Campaign within the U.S. Department of Energy smore » Office of Nuclear Energy. The approach in this paper identifies the features needed to cover the range of promising fuel cycle options identified in the DOE-NE Fuel Cycle Evaluation and Screening (E&S) and categorizes these features to facilitate prioritization. Features were categorized as essential functions, integrating features, and exemplary capabilities. One objective of this paper is to propose a library of unit tests applicable to each of the essential functions. Another underlying motivation for this paper is to encourage an international dialog on the functionalities and standard test methods for fuel cycle simulator tools.« less

Temperature measuring analysis of the nuclear reactor fuel assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Urban, F., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Kučák, L., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Bereznai, J., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk

2014-08-06

Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuelmore » assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.« less

Current conducting end plate of fuel cell assembly

DOEpatents

Walsh, Michael M.

1999-01-01

A fuel cell assembly has a current conducting end plate with a conductive body formed integrally with isolating material. The conductive body has a first surface, a second surface opposite the first surface, and an electrical connector. The first surface has an exposed portion for conducting current between a working section of the fuel cell assembly and the electrical connector. The isolating material is positioned on at least a portion of the second surface. The conductive body can have support passage(s) extending therethrough for receiving structural member(s) of the fuel cell assembly. Isolating material can electrically isolate the conductive body from the structural member(s). The conductive body can have service passage(s) extending therethrough for servicing one or more fluids for the fuel cell assembly. Isolating material can chemically isolate the one or more fluids from the conductive body. The isolating material can also electrically isolate the conductive body from the one or more fluids.

Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric Richard; Durbin, Samuel G

2007-04-01

The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program providedmore » data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.« less

Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

DOEpatents

Oosterkamp, Willem Jan; Marquino, Wayne

1999-01-05

A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

PWR and BWR spent fuel assembly gamma spectra measurements

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

2016-10-01

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

PWR and BWR spent fuel assembly gamma spectra measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

PWR and BWR spent fuel assembly gamma spectra measurements

DOE PAGES

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

2016-07-17

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

Identification of fuel cycle simulator functionalities for analysis of transition to a new fuel cycle

DOE PAGES

Brown, Nicholas R.; Carlsen, Brett W.; Dixon, Brent W.; ...

2016-06-09

Dynamic fuel cycle simulation tools are intended to model holistic transient nuclear fuel cycle scenarios. As with all simulation tools, fuel cycle simulators require verification through unit tests, benchmark cases, and integral tests. Model validation is a vital aspect as well. Although compara-tive studies have been performed, there is no comprehensive unit test and benchmark library for fuel cycle simulator tools. The objective of this paper is to identify the must test functionalities of a fuel cycle simulator tool within the context of specific problems of interest to the Fuel Cycle Options Campaign within the U.S. Department of Energy smore » Office of Nuclear Energy. The approach in this paper identifies the features needed to cover the range of promising fuel cycle options identified in the DOE-NE Fuel Cycle Evaluation and Screening (E&S) and categorizes these features to facilitate prioritization. Features were categorized as essential functions, integrating features, and exemplary capabilities. One objective of this paper is to propose a library of unit tests applicable to each of the essential functions. Another underlying motivation for this paper is to encourage an international dialog on the functionalities and standard test methods for fuel cycle simulator tools.« less

Critical Assessment of Metagenome Interpretation – a benchmark of computational metagenomics software

PubMed Central

Sczyrba, Alexander; Hofmann, Peter; Belmann, Peter; Koslicki, David; Janssen, Stefan; Dröge, Johannes; Gregor, Ivan; Majda, Stephan; Fiedler, Jessika; Dahms, Eik; Bremges, Andreas; Fritz, Adrian; Garrido-Oter, Ruben; Jørgensen, Tue Sparholt; Shapiro, Nicole; Blood, Philip D.; Gurevich, Alexey; Bai, Yang; Turaev, Dmitrij; DeMaere, Matthew Z.; Chikhi, Rayan; Nagarajan, Niranjan; Quince, Christopher; Meyer, Fernando; Balvočiūtė, Monika; Hansen, Lars Hestbjerg; Sørensen, Søren J.; Chia, Burton K. H.; Denis, Bertrand; Froula, Jeff L.; Wang, Zhong; Egan, Robert; Kang, Dongwan Don; Cook, Jeffrey J.; Deltel, Charles; Beckstette, Michael; Lemaitre, Claire; Peterlongo, Pierre; Rizk, Guillaume; Lavenier, Dominique; Wu, Yu-Wei; Singer, Steven W.; Jain, Chirag; Strous, Marc; Klingenberg, Heiner; Meinicke, Peter; Barton, Michael; Lingner, Thomas; Lin, Hsin-Hung; Liao, Yu-Chieh; Silva, Genivaldo Gueiros Z.; Cuevas, Daniel A.; Edwards, Robert A.; Saha, Surya; Piro, Vitor C.; Renard, Bernhard Y.; Pop, Mihai; Klenk, Hans-Peter; Göker, Markus; Kyrpides, Nikos C.; Woyke, Tanja; Vorholt, Julia A.; Schulze-Lefert, Paul; Rubin, Edward M.; Darling, Aaron E.; Rattei, Thomas; McHardy, Alice C.

2018-01-01

In metagenome analysis, computational methods for assembly, taxonomic profiling and binning are key components facilitating downstream biological data interpretation. However, a lack of consensus about benchmarking datasets and evaluation metrics complicates proper performance assessment. The Critical Assessment of Metagenome Interpretation (CAMI) challenge has engaged the global developer community to benchmark their programs on datasets of unprecedented complexity and realism. Benchmark metagenomes were generated from ~700 newly sequenced microorganisms and ~600 novel viruses and plasmids, including genomes with varying degrees of relatedness to each other and to publicly available ones and representing common experimental setups. Across all datasets, assembly and genome binning programs performed well for species represented by individual genomes, while performance was substantially affected by the presence of related strains. Taxonomic profiling and binning programs were proficient at high taxonomic ranks, with a notable performance decrease below the family level. Parameter settings substantially impacted performances, underscoring the importance of program reproducibility. While highlighting current challenges in computational metagenomics, the CAMI results provide a roadmap for software selection to answer specific research questions. PMID:28967888

Burnable absorber arrangement for fuel bundle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crowther, R.L.; Townsend, D.B.

1986-12-16

This patent describes a boiling water reactor core whose operation is characterized by a substantial proportion of steam voids with concomitantly reduced moderation toward the top of the core when the reactor is in its hot operating condition. The reduced moderation leads to slower burnup and greater conversion ratio in an upper core region so that when the reactor is in its cold shut down condition the resulting relatively increased moderation in the upper core region is accompanied by a reactivity profile that peaks in the upper core region. A fuel assembly is described comprising; a component of fissile materialmore » distributed over a substantial axial extent of the fuel assembly; and a component of neutron absorbing material having an axial distribution characterized by an enhancement in an axial zone of the fuel assembly, designated the cold shutdown control zone, corresponding to at least a portion of the axial region of the core when the cold shutdown reactivity peaks. The aggregate amount of neutron absorbing material in the cold shutdown zone of the fuel assembly is greater than the aggregate amount of neutron absorbing material in the axial zones of the fuel assembly immediately above and immediately below the cold shutdown control zone whereby the cold shutdown reactivity peak is reduced relative to the cold shutdown reactivity in the zones immediately above and immediately below the cold shutdown control zone. The cold shutdown zone has an axial extent measured from the bottom of the fuel assembly in the range between 68-88 percent of the height of the fissile material in the fuel assembly.« less

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

2016-05-01

This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

2016-05-12

This report describes the third set of tests (the “DCL a shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

DOEpatents

Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ

2009-12-22

The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

Fuel cell design and assembly

DOEpatents

Myerhoff, Alfred

1984-01-01

The present invention is directed to a novel bipolar cooling plate, fuel cell design and method of assembly of fuel cells. The bipolar cooling plate used in the fuel cell design and method of assembly has discrete opposite edge and means carried by the plate defining a plurality of channels extending along the surface of the plate toward the opposite edges. At least one edge of the channels terminates short of the edge of the plate defining a recess for receiving a fastener.

Advanced techniques for repair of irradiated PWR fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knaab, H.; Westphal, M.

Kraftwerk union has recently designed and built a portable repair unit for use in nuclear power plants for repair of defective fuel assemblies where space limitations do not allow permanent installation of repair equipment. This new equipment is designed to be easily disassembled and decontaminated. The main component of the equipment is the fuel assembly reconstitution unit (FARU) which is placed on the floor of the spent fuel pool. The use of the FARU is described in the paper.

Fuel nozzle assembly

DOEpatents

Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

2011-08-30

A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

Benchmark Evaluation of the HTR-PROTEUS Absorber Rod Worths (Core 4)

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; Leland M. Montierth

2014-06-01

PROTEUS was a zero-power research reactor at the Paul Scherrer Institute (PSI) in Switzerland. The critical assembly was constructed from a large graphite annulus surrounding a central cylindrical cavity. Various experimental programs were investigated in PROTEUS; during the years 1992 through 1996, it was configured as a pebble-bed reactor and designated HTR-PROTEUS. Various critical configurations were assembled with each accompanied by an assortment of reactor physics experiments including differential and integral absorber rod measurements, kinetics, reaction rate distributions, water ingress effects, and small sample reactivity effects [1]. Four benchmark reports were previously prepared and included in the March 2013 editionmore » of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [2] evaluating eleven critical configurations. A summary of that effort was previously provided [3] and an analysis of absorber rod worth measurements for Cores 9 and 10 have been performed prior to this analysis and included in PROTEUS-GCR-EXP-004 [4]. In the current benchmark effort, absorber rod worths measured for Core Configuration 4, which was the only core with a randomly-packed pebble loading, have been evaluated for inclusion as a revision to the HTR-PROTEUS benchmark report PROTEUS-GCR-EXP-002.« less

Internal reforming fuel cell assembly with simplified fuel feed

DOEpatents

Farooque, Mohammad; Novacco, Lawrence J.; Allen, Jeffrey P.

2001-01-01

A fuel cell assembly in which fuel cells adapted to internally reform fuel and fuel reformers for reforming fuel are arranged in a fuel cell stack. The fuel inlet ports of the fuel cells and the fuel inlet ports and reformed fuel outlet ports of the fuel reformers are arranged on one face of the fuel cell stack. A manifold sealing encloses this face of the stack and a reformer fuel delivery system is arranged entirely within the region between the manifold and the one face of the stack. The fuel reformer has a foil wrapping and a cover member forming with the foil wrapping an enclosed structure.

Evaluation of RAPID for a UNF cask benchmark problem

NASA Astrophysics Data System (ADS)

Mascolino, Valerio; Haghighat, Alireza; Roskoff, Nathan J.

2017-09-01

This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system for the simulation of a used nuclear fuel (UNF) cask. RAPID is capable of determining eigenvalue, subcritical multiplication, and pin-wise, axially-dependent fission density throughout a UNF cask. We study the source convergence based on the analysis of the different parameters used in an eigenvalue calculation in the MCNP Monte Carlo code. For this study, we consider a single assembly surrounded by absorbing plates with reflective boundary conditions. Based on the best combination of eigenvalue parameters, a reference MCNP solution for the single assembly is obtained. RAPID results are in excellent agreement with the reference MCNP solutions, while requiring significantly less computation time (i.e., minutes vs. days). A similar set of eigenvalue parameters is used to obtain a reference MCNP solution for the whole UNF cask. Because of time limitation, the MCNP results near the cask boundaries have significant uncertainties. Except for these, the RAPID results are in excellent agreement with the MCNP predictions, and its computation time is significantly lower, 35 second on 1 core versus 9.5 days on 16 cores.

Combustor assembly for use in a turbine engine and methods of assembling same

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward

2013-05-14

A fuel nozzle assembly for use with a turbine engine is described herein. The fuel nozzle assembly includes a plurality of fuel nozzles positioned within an air plenum defined by a casing. Each of the plurality of fuel nozzles is coupled to a combustion liner defining a combustion chamber. Each of the plurality of fuel nozzles includes a housing that includes an inner surface that defines a cooling fluid plenum and a fuel plenum therein, and a plurality of mixing tubes extending through the housing. Each of the mixing tubes includes an inner surface defining a flow channel extending between the air plenum and the combustion chamber. At least one mixing tube of the plurality of mixing tubes including at least one cooling fluid aperture for channeling a flow of cooling fluid from the cooling fluid plenum to the flow channel.

Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

DOEpatents

Phillips, John R.; Halbig, James K.; Menlove, Howard O.; Klosterbuer, Shirley F.

1985-01-01

A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

DOEpatents

Oosterkamp, W.J.; Marquino, W.

1999-01-05

A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies is disclosed. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereas access to the fuel assemblies is not obstructed. 11 figs.

Hybrid Gama Emission Tomography (HGET): FY16 Annual Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Erin A.; Smith, Leon E.; Wittman, Richard S.

2017-02-01

Current International Atomic Energy Agency (IAEA) methodologies for the verification of fresh low-enriched uranium (LEU) and mixed oxide (MOX) fuel assemblies are volume-averaging methods that lack sensitivity to individual pins. Further, as fresh fuel assemblies become more and more complex (e.g., heavy gadolinium loading, high degrees of axial and radial variation in fissile concentration), the accuracy of current IAEA instruments degrades and measurement time increases. Particularly in light of the fact that no special tooling is required to remove individual pins from modern fuel assemblies, the IAEA needs new capabilities for the verification of unirradiated (i.e., fresh LEU and MOX)more » assemblies to ensure that fissile material has not been diverted. Passive gamma emission tomography has demonstrated potential to provide pin-level verification of spent fuel, but gamma-ray emission rates from unirradiated fuel emissions are significantly lower, precluding purely passive tomography methods. The work presented here introduces the concept of Hybrid Gamma Emission Tomography (HGET) for verification of unirradiated fuels, in which a neutron source is used to actively interrogate the fuel assembly and the resulting gamma-ray emissions are imaged using tomographic methods to provide pin-level verification of fissile material concentration.« less

Fuel-Mediated Transient Clustering of Colloidal Building Blocks.

PubMed

van Ravensteijn, Bas G P; Hendriksen, Wouter E; Eelkema, Rienk; van Esch, Jan H; Kegel, Willem K

2017-07-26

Fuel-driven assembly operates under the continuous influx of energy and results in superstructures that exist out of equilibrium. Such dissipative processes provide a route toward structures and transient behavior unreachable by conventional equilibrium self-assembly. Although perfected in biological systems like microtubules, this class of assembly is only sparsely used in synthetic or colloidal analogues. Here, we present a novel colloidal system that shows transient clustering driven by a chemical fuel. Addition of fuel causes an increase in hydrophobicity of the building blocks by actively removing surface charges, thereby driving their aggregation. Depletion of fuel causes reappearance of the charged moieties and leads to disassembly of the formed clusters. This reassures that the system returns to its initial, equilibrium state. By taking advantage of the cyclic nature of our system, we show that clustering can be induced several times by simple injection of new fuel. The fuel-mediated assembly of colloidal building blocks presented here opens new avenues to the complex landscape of nonequilibrium colloidal structures, guided by biological design principles.

SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

DOEpatents

Schluderberg, D.C.; Ryon, J.W.

1962-05-01

A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

Air breathing direct methanol fuel cell

DOEpatents

Ren, Xiaoming; Gottesfeld, Shimshon

2002-01-01

An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol source. Water loss from the cell is minimized by making the conductive cathode assembly hydrophobic and the conductive anode assembly hydrophilic.

Nuclear imaging of the fuel assembly in ignition experimentsa)

NASA Astrophysics Data System (ADS)

Grim, G. P.; Guler, N.; Merrill, F. E.; Morgan, G. L.; Danly, C. R.; Volegov, P. L.; Wilde, C. H.; Wilson, D. C.; Clark, D. S.; Hinkel, D. E.; Jones, O. S.; Raman, K. S.; Izumi, N.; Fittinghoff, D. N.; Drury, O. B.; Alger, E. T.; Arnold, P. A.; Ashabranner, R. C.; Atherton, L. J.; Barrios, M. A.; Batha, S.; Bell, P. M.; Benedetti, L. R.; Berger, R. L.; Bernstein, L. A.; Berzins, L. V.; Betti, R.; Bhandarkar, S. D.; Bionta, R. M.; Bleuel, D. L.; Boehly, T. R.; Bond, E. J.; Bowers, M. W.; Bradley, D. K.; Brunton, G. K.; Buckles, R. A.; Burkhart, S. C.; Burr, R. F.; Caggiano, J. A.; Callahan, D. A.; Casey, D. T.; Castro, C.; Celliers, P. M.; Cerjan, C. J.; Chandler, G. A.; Choate, C.; Cohen, S. J.; Collins, G. W.; Cooper, G. W.; Cox, J. R.; Cradick, J. R.; Datte, P. S.; Dewald, E. L.; Di Nicola, P.; Di Nicola, J. M.; Divol, L.; Dixit, S. N.; Dylla-Spears, R.; Dzenitis, E. G.; Eckart, M. J.; Eder, D. C.; Edgell, D. H.; Edwards, M. J.; Eggert, J. H.; Ehrlich, R. B.; Erbert, G. V.; Fair, J.; Farley, D. R.; Felker, B.; Fortner, R. J.; Frenje, J. A.; Frieders, G.; Friedrich, S.; Gatu-Johnson, M.; Gibson, C. R.; Giraldez, E.; Glebov, V. Y.; Glenn, S. M.; Glenzer, S. H.; Gururangan, G.; Haan, S. W.; Hahn, K. D.; Hammel, B. A.; Hamza, A. V.; Hartouni, E. P.; Hatarik, R.; Hatchett, S. P.; Haynam, C.; Hermann, M. R.; Herrmann, H. W.; Hicks, D. G.; Holder, J. P.; Holunga, D. M.; Horner, J. B.; Hsing, W. W.; Huang, H.; Jackson, M. C.; Jancaitis, K. S.; Kalantar, D. H.; Kauffman, R. L.; Kauffman, M. I.; Khan, S. F.; Kilkenny, J. D.; Kimbrough, J. R.; Kirkwood, R.; Kline, J. L.; Knauer, J. P.; Knittel, K. M.; Koch, J. A.; Kohut, T. R.; Kozioziemski, B. J.; Krauter, K.; Krauter, G. W.; Kritcher, A. L.; Kroll, J.; Kyrala, G. A.; Fortune, K. N. La; LaCaille, G.; Lagin, L. J.; Land, T. A.; Landen, O. L.; Larson, D. W.; Latray, D. A.; Leeper, R. J.; Lewis, T. L.; LePape, S.; Lindl, J. D.; Lowe-Webb, R. R.; Ma, T.; MacGowan, B. J.; MacKinnon, A. J.; MacPhee, A. G.; Malone, R. M.; Malsbury, T. N.; Mapoles, E.; Marshall, C. D.; Mathisen, D. G.; McKenty, P.; McNaney, J. M.; Meezan, N. B.; Michel, P.; Milovich, J. L.; Moody, J. D.; Moore, A. S.; Moran, M. J.; Moreno, K.; Moses, E. I.; Munro, D. H.; Nathan, B. R.; Nelson, A. J.; Nikroo, A.; Olson, R. E.; Orth, C.; Pak, A. E.; Palma, E. S.; Parham, T. G.; Patel, P. K.; Patterson, R. W.; Petrasso, R. D.; Prasad, R.; Ralph, J. E.; Regan, S. P.; Rinderknecht, H.; Robey, H. F.; Ross, G. F.; Ruiz, C. L.; Séguin, F. H.; Salmonson, J. D.; Sangster, T. C.; Sater, J. D.; Saunders, R. L.; Schneider, M. B.; Schneider, D. H.; Shaw, M. J.; Simanovskaia, N.; Spears, B. K.; Springer, P. T.; Stoeckl, C.; Stoeffl, W.; Suter, L. J.; Thomas, C. A.; Tommasini, R.; Town, R. P.; Traille, A. J.; Wonterghem, B. Van; Wallace, R. J.; Weaver, S.; Weber, S. V.; Wegner, P. J.; Whitman, P. K.; Widmann, K.; Widmayer, C. C.; Wood, R. D.; Young, B. K.; Zacharias, R. A.; Zylstra, A.

2013-05-01

First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium-tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models' prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%-25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

Upgraded HFIR Fuel Element Welding System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sease, John D

2010-02-01

The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. Inmore » recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.« less

A User’s Guide to the PLTEMP/ANL Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, A. P.; Kalimullah, M.; Feldman, E. E.

2016-07-25

PLTEMP/ANL V4.2 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each withmore » its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.« less

Development and verification of NRC`s single-rod fuel performance codes FRAPCON-3 AND FRAPTRAN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beyer, C.E.; Cunningham, M.E.; Lanning, D.D.

1998-03-01

The FRAPCON and FRAP-T code series, developed in the 1970s and early 1980s, are used by the US Nuclear Regulatory Commission (NRC) to predict fuel performance during steady-state and transient power conditions, respectively. Both code series are now being updated by Pacific Northwest National Laboratory to improve their predictive capabilities at high burnup levels. The newest versions of the codes are called FRAPCON-3 and FRAPTRAN. The updates to fuel property and behavior models are focusing on providing best estimate predictions under steady-state and fast transient power conditions up to extended fuel burnups (> 55 GWd/MTU). Both codes will be assessedmore » against a data base independent of the data base used for code benchmarking and an estimate of code predictive uncertainties will be made based on comparisons to the benchmark and independent data bases.« less

75 FR 901 - Airworthiness Directives; Honeywell International Inc. ALF502 Series and LF507 Series Turbofan...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-07

.... ALF502 series and LF507 series turbofan engines with certain fuel manifold assemblies installed. That AD... of certain part number (P/N) fuel manifold assemblies for cracks, and replacement of cracked fuel manifolds with serviceable manifolds. This AD continues to require inspecting those fuel manifolds for...

77 FR 72200 - Airworthiness Directives; The Boeing Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-05

... correctly on the engine fuel feed manifold couplings. This AD also requires inspecting the assembly of the engine fuel feed manifold rigid and full flexible couplings. This AD was prompted by reports of fuel leaks due to improperly assembled engine fuel feed manifold couplings. We are issuing this AD to detect...

Fuel cell cooler-humidifier plate

DOEpatents

Vitale, Nicholas G.; Jones, Daniel O.

2000-01-01

A cooler-humidifier plate for use in a proton exchange membrane (PEM) fuel cell stack assembly is provided. The cooler-humidifier plate combines functions of cooling and humidification within the fuel cell stack assembly, thereby providing a more compact structure, simpler manifolding, and reduced reject heat from the fuel cell. Coolant on the cooler side of the plate removes heat generated within the fuel cell assembly. Heat is also removed by the humidifier side of the plate for use in evaporating the humidification water. On the humidifier side of the plate, evaporating water humidifies reactant gas flowing over a moistened wick. After exiting the humidifier side of the plate, humidified reactant gas provides needed moisture to the proton exchange membranes used in the fuel cell stack assembly. The invention also provides a fuel cell plate that maximizes structural support within the fuel cell by ensuring that the ribs that form the boundaries of channels on one side of the plate have ends at locations that substantially correspond to the locations of ribs on the opposite side of the plate.

Automatically closing swing gate closure assembly

DOEpatents

Chang, Shih-Chih; Schuck, William J.; Gilmore, Richard F.

1988-01-01

A swing gate closure assembly for nuclear reactor tipoff assembly wherein the swing gate is cammed open by a fuel element or spacer but is reliably closed at a desired closing rate primarily by hydraulic forces in the absence of a fuel charge.

End plate assembly having a two-phase fluid-filled bladder and method for compressing a fuel cell stack

DOEpatents

Carlstrom, Jr., Charles M.

2001-01-01

An end plate assembly is disclosed for use in a fuel cell assembly in which the end plate assembly includes a housing having a cavity, and a bladder receivable in the cavity and engageable with the fuel cell stack. The bladder includes a two-phase fluid having a liquid portion and a vapor portion. Desirably, the two-phase fluid has a vapor pressure between about 100 psi and about 600 psi at a temperature between about 70 degrees C. to about 110 degrees C.

Nuclear data uncertainty propagation by the XSUSA method in the HELIOS2 lattice code

NASA Astrophysics Data System (ADS)

Wemple, Charles; Zwermann, Winfried

2017-09-01

Uncertainty quantification has been extensively applied to nuclear criticality analyses for many years and has recently begun to be applied to depletion calculations. However, regulatory bodies worldwide are trending toward requiring such analyses for reactor fuel cycle calculations, which also requires uncertainty propagation for isotopics and nuclear reaction rates. XSUSA is a proven methodology for cross section uncertainty propagation based on random sampling of the nuclear data according to covariance data in multi-group representation; HELIOS2 is a lattice code widely used for commercial and research reactor fuel cycle calculations. This work describes a technique to automatically propagate the nuclear data uncertainties via the XSUSA approach through fuel lattice calculations in HELIOS2. Application of the XSUSA methodology in HELIOS2 presented some unusual challenges because of the highly-processed multi-group cross section data used in commercial lattice codes. Currently, uncertainties based on the SCALE 6.1 covariance data file are being used, but the implementation can be adapted to other covariance data in multi-group structure. Pin-cell and assembly depletion calculations, based on models described in the UAM-LWR Phase I and II benchmarks, are performed and uncertainties in multiplication factor, reaction rates, isotope concentrations, and delayed-neutron data are calculated. With this extension, it will be possible for HELIOS2 users to propagate nuclear data uncertainties directly from the microscopic cross sections to subsequent core simulations.

Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir

2014-09-18

Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performedmore » in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.« less

Dependency of the Reynolds number on the water flow through the perforated tube

DOE Office of Scientific and Technical Information (OSTI.GOV)

Závodný, Zdenko, E-mail: zdenko.zavodny@stuba.sk; Bereznai, Jozef, E-mail: jozef.bereznai@stuba.sk; Urban, František

Safe and effective loading of nuclear reactor fuel assemblies demands qualitative and quantitative analysis of the relationship between the coolant temperature in the fuel assembly outlet, measured by the thermocouple, and the mean coolant temperature profile in the thermocouple plane position. It is not possible to perform the analysis directly in the reactor, so it is carried out using measurements on the physical model, and the CFD fuel assembly coolant flow models. The CFD models have to be verified and validated in line with the temperature and velocity profile obtained from the measurements of the cooling water flowing in themore » physical model of the fuel assembly. Simplified physical model with perforated central tube and its validated CFD model serve to design of the second physical model of the fuel assembly of the nuclear reactor VVER 440. Physical model will be manufactured and installed in the laboratory of the Institute of Energy Machines, Faculty of Mechanical Engineering of the Slovak University of Technology in Bratislava.« less

Miniature ceramic fuel cell

DOEpatents

Lessing, Paul A.; Zuppero, Anthony C.

1997-06-24

A miniature power source assembly capable of providing portable electricity is provided. A preferred embodiment of the power source assembly employing a fuel tank, fuel pump and control, air pump, heat management system, power chamber, power conditioning and power storage. The power chamber utilizes a ceramic fuel cell to produce the electricity. Incoming hydro carbon fuel is automatically reformed within the power chamber. Electrochemical combustion of hydrogen then produces electricity.

78 FR 76328 - Application for Renewal of Special Nuclear Material License SNM-2014 From Tennessee Valley...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-17

... SNM in the form of fully-assembled fuel assemblies that would later form the initial reactor core of WBN2. The SNM in the fuel assemblies is enriched up to 5% in the isotope U-235. The fresh fuel... received the initial core for WBN2. The NRC has not yet issued the OL for the Unit 2 reactor. The...

Teledyne Energy Systems, Inc., Proton Exchange Member (PEM) Fuel Cell Engineering Model Powerplant. Test Report: Initial Benchmark Tests in the Original Orientation

NASA Technical Reports Server (NTRS)

Loyselle, Patricia; Prokopius, Kevin

2011-01-01

Proton Exchange Membrane (PEM) fuel cell technology is the leading candidate to replace the alkaline fuel cell technology, currently used on the Shuttle, for future space missions. During a 5-yr development program, a PEM fuel cell powerplant was developed. This report details the initial performance evaluation test results of the powerplant.

Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay,more » and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.« less

Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

NASA Astrophysics Data System (ADS)

Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.; Menlove, Howard O.; Flaska, Marek; Pozzi, Sara A.

2017-07-01

The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrument that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.

Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.

The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. Thus the NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrumentmore » that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.« less

Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

DOE PAGES

Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.; ...

2017-01-05

The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. Thus the NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrumentmore » that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.« less

METHOD AND MEANS FOR SUPPORTING REACTOR FUEL CONTAINERS IN AN ASSEMBLY

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.; Coombs, C.A.

1962-12-11

This patent relates to means for supporting fuelcontaining tubes in an assembly which include grid means at either end of the fuel element assembly antl improved grid means intermediate of the ends to provide support against lateral displacement. (AEC)

Solid oxide fuel cell generator with removable modular fuel cell stack configurations

DOEpatents

Gillett, J.E.; Dederer, J.T.; Zafred, P.R.; Collie, J.C.

1998-04-21

A high temperature solid oxide fuel cell generator produces electrical power from oxidation of hydrocarbon fuel gases such as natural gas, or conditioned fuel gases, such as carbon monoxide or hydrogen, with oxidant gases, such as air or oxygen. This electrochemical reaction occurs in a plurality of electrically connected solid oxide fuel cells bundled and arrayed in a unitary modular fuel cell stack disposed in a compartment in the generator container. The use of a unitary modular fuel cell stack in a generator is similar in concept to that of a removable battery. The fuel cell stack is provided in a pre-assembled self-supporting configuration where the fuel cells are mounted to a common structural base having surrounding side walls defining a chamber. Associated generator equipment may also be mounted to the fuel cell stack configuration to be integral therewith, such as a fuel and oxidant supply and distribution systems, fuel reformation systems, fuel cell support systems, combustion, exhaust and spent fuel recirculation systems, and the like. The pre-assembled self-supporting fuel cell stack arrangement allows for easier assembly, installation, maintenance, better structural support and longer life of the fuel cells contained in the fuel cell stack. 8 figs.

Solid oxide fuel cell generator with removable modular fuel cell stack configurations

DOEpatents

Gillett, James E.; Dederer, Jeffrey T.; Zafred, Paolo R.; Collie, Jeffrey C.

1998-01-01

A high temperature solid oxide fuel cell generator produces electrical power from oxidation of hydrocarbon fuel gases such as natural gas, or conditioned fuel gases, such as carbon monoxide or hydrogen, with oxidant gases, such as air or oxygen. This electrochemical reaction occurs in a plurality of electrically connected solid oxide fuel cells bundled and arrayed in a unitary modular fuel cell stack disposed in a compartment in the generator container. The use of a unitary modular fuel cell stack in a generator is similar in concept to that of a removable battery. The fuel cell stack is provided in a pre-assembled self-supporting configuration where the fuel cells are mounted to a common structural base having surrounding side walls defining a chamber. Associated generator equipment may also be mounted to the fuel cell stack configuration to be integral therewith, such as a fuel and oxidant supply and distribution systems, fuel reformation systems, fuel cell support systems, combustion, exhaust and spent fuel recirculation systems, and the like. The pre-assembled self-supporting fuel cell stack arrangement allows for easier assembly, installation, maintenance, better structural support and longer life of the fuel cells contained in the fuel cell stack.

Separating homeologs by phasing in the tetraploid wheat transcriptome.

PubMed

Krasileva, Ksenia V; Buffalo, Vince; Bailey, Paul; Pearce, Stephen; Ayling, Sarah; Tabbita, Facundo; Soria, Marcelo; Wang, Shichen; Akhunov, Eduard; Uauy, Cristobal; Dubcovsky, Jorge

2013-06-25

The high level of identity among duplicated homoeologous genomes in tetraploid pasta wheat presents substantial challenges for de novo transcriptome assembly. To solve this problem, we develop a specialized bioinformatics workflow that optimizes transcriptome assembly and separation of merged homoeologs. To evaluate our strategy, we sequence and assemble the transcriptome of one of the diploid ancestors of pasta wheat, and compare both assemblies with a benchmark set of 13,472 full-length, non-redundant bread wheat cDNAs. A total of 489 million 100 bp paired-end reads from tetraploid wheat assemble in 140,118 contigs, including 96% of the benchmark cDNAs. We used a comparative genomics approach to annotate 66,633 open reading frames. The multiple k-mer assembly strategy increases the proportion of cDNAs assembled full-length in a single contig by 22% relative to the best single k-mer size. Homoeologs are separated using a post-assembly pipeline that includes polymorphism identification, phasing of SNPs, read sorting, and re-assembly of phased reads. Using a reference set of genes, we determine that 98.7% of SNPs analyzed are correctly separated by phasing. Our study shows that de novo transcriptome assembly of tetraploid wheat benefit from multiple k-mer assembly strategies more than diploid wheat. Our results also demonstrate that phasing approaches originally designed for heterozygous diploid organisms can be used to separate the close homoeologous genomes of tetraploid wheat. The predicted tetraploid wheat proteome and gene models provide a valuable tool for the wheat research community and for those interested in comparative genomic studies.

Separating homeologs by phasing in the tetraploid wheat transcriptome

PubMed Central

2013-01-01

Background The high level of identity among duplicated homoeologous genomes in tetraploid pasta wheat presents substantial challenges for de novo transcriptome assembly. To solve this problem, we develop a specialized bioinformatics workflow that optimizes transcriptome assembly and separation of merged homoeologs. To evaluate our strategy, we sequence and assemble the transcriptome of one of the diploid ancestors of pasta wheat, and compare both assemblies with a benchmark set of 13,472 full-length, non-redundant bread wheat cDNAs. Results A total of 489 million 100 bp paired-end reads from tetraploid wheat assemble in 140,118 contigs, including 96% of the benchmark cDNAs. We used a comparative genomics approach to annotate 66,633 open reading frames. The multiple k-mer assembly strategy increases the proportion of cDNAs assembled full-length in a single contig by 22% relative to the best single k-mer size. Homoeologs are separated using a post-assembly pipeline that includes polymorphism identification, phasing of SNPs, read sorting, and re-assembly of phased reads. Using a reference set of genes, we determine that 98.7% of SNPs analyzed are correctly separated by phasing. Conclusions Our study shows that de novo transcriptome assembly of tetraploid wheat benefit from multiple k-mer assembly strategies more than diploid wheat. Our results also demonstrate that phasing approaches originally designed for heterozygous diploid organisms can be used to separate the close homoeologous genomes of tetraploid wheat. The predicted tetraploid wheat proteome and gene models provide a valuable tool for the wheat research community and for those interested in comparative genomic studies. PMID:23800085

Benchmarking criticality analysis of TRIGA fuel storage racks.

PubMed

Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

2017-01-01

A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel. Copyright © 2016. Published by Elsevier Ltd.

96. SEED 1 FUEL ASSEMBLY FROM LOCATION L9 BEING REMOVED ...

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

96. SEED 1 FUEL ASSEMBLY FROM LOCATION L-9 BEING REMOVED FROM REACTOR VESSEL BY MEANS OF FUEL EXTRACTION CRANE, JANUARY 7, 1960 - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP603) SHOWING CRANE ASSEMBLY ...

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

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP-603) SHOWING CRANE ASSEMBLY FOR TRANSFER PIT. INL PHOTO NUMBER NRTS-51-2404. Unknown Photographer, 5/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Nuclear reactor composite fuel assembly

DOEpatents

Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

1980-01-01

A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

Passive Safety Features Evaluation of KIPT Neutron Source Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhong, Zhaopeng; Gohar, Yousry

2016-06-01

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. It will be used to conduct basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The facility has an electron accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assemblymore » is composed of WWR-M2 - Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The subcritical assembly is seated in a water tank, which is a part of the primary cooling loop. During normal operation, the water coolant operates at room temperature and the total facility power is ~300 KW. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/B-VII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight water temperature increase and the corresponding water density decrease produce large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. The reactivity variation due to an empty fuel position filled by water during the fuel loading process is examined. Also, the loading mistakes of removing beryllium reflector assemblies and replacing them with dummy assemblies were analyzed. In all these circumstances, the reactivity change results do not cause any safety concerns.« less

Air breathing direct methanol fuel cell

DOEpatents

Ren, Xiaoming

2002-01-01

An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol source.

Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

2012-07-01

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-11-07

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-01-01

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

Delayed Gamma-ray Spectroscopy for Safeguards Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mozin, Vladimir

The delayed gamma-ray assay technique utilizes an external neutron source (D-D, D-T, or electron accelerator-driven), and high-resolution gamma-ray spectrometers to perform characterization of SNM materials behind shielding and in complex configurations such as a nuclear fuel assembly. High-energy delayed gamma-rays (2.5 MeV and above) observed following the active interrogation, provide a signature for identification of specific fissionable isotopes in a mixed sample, and determine their relative content. Potential safeguards applications of this method are: 1) characterization of fresh and spent nuclear fuel assemblies in wet or dry storage; 2) analysis of uranium enrichment in shielded or non-characterized containers or inmore » the presence of a strong radioactive background and plutonium contamination; 3) characterization of bulk and waste and product streams at SNM processing plants. Extended applications can include warhead confirmation and warhead dismantlement confirmation in the arms control area, as well as SNM diagnostics for the emergency response needs. In FY16 and prior years, the project has demonstrated the delayed gamma-ray measurement technique as a robust SNM assay concept. A series of empirical and modeling studies were conducted to characterize its response sensitivity, develop analysis methodologies, and analyze applications. Extensive experimental tests involving weapons-grade Pu, HEU and depleted uranium samples were completed at the Idaho Accelerator Center and LLNL Dome facilities for various interrogation time regimes and effects of the neutron source parameters. A dedicated delayed gamma-ray response modeling technique was developed and its elements were benchmarked in representative experimental studies, including highresolution gamma-ray measurements of spent fuel at the CLAB facility in Sweden. The objective of the R&D effort in FY17 is to experimentally demonstrate the feasibility of the delayed gamma-ray interrogation of shielded SNM samples with portable neutron sources suitable for field applications.« less

Square lattice honeycomb reactor for space power and propulsion

NASA Astrophysics Data System (ADS)

Gouw, Reza; Anghaie, Samim

2000-01-01

The most recent nuclear design study at the Innovative Nuclear Space Power and Propulsion Institute (INSPI) is the Moderated Square-Lattice Honeycomb (M-SLHC) reactor design utilizing the solid solution of ternary carbide fuels. The reactor is fueled with solid solution of 93% enriched (U,Zr,Nb)C. The square-lattice honeycomb design provides high strength and is amenable to the processing complexities of these ultrahigh temperature fuels. The optimum core configuration requires a balance between high specific impulse and thrust level performance, and maintaining the temperature and strength limits of the fuel. The M-SLHC design is based on a cylindrical core that has critical radius and length of 37 cm and 50 cm, respectively. This design utilized zirconium hydrate to act as moderator. The fuel sub-assemblies are designed as cylindrical tubes with 12 cm in diameter and 10 cm in length. Five fuel subassemblies are stacked up axially to form one complete fuel assembly. These fuel assemblies are then arranged in the circular arrangement to form two fuel regions. The first fuel region consists of six fuel assemblies, and 18 fuel assemblies for the second fuel region. A 10-cm radial beryllium reflector in addition to 10-cm top axial beryllium reflector is used to reduce neutron leakage from the system. To perform nuclear design analysis of the M-SLHC design, a series of neutron transport and diffusion codes are used. To optimize the system design, five axial regions are specified. In each axial region, temperature and fuel density are varied. The axial and radial power distributions for the system are calculated, as well as the axial and radial flux distributions. Temperature coefficients of the system are also calculated. A water submersion accident scenario is also analyzed for these systems. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel, which provides a relatively high thrust to weight ratio. .

Nuclear imaging of the fuel assembly in ignition experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grim, G. P.; Guler, N.; Merrill, F. E.

First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate thatmore » the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models’ prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.« less

Analytic Methods for Benchmarking Hydrogen and Fuel Cell Technologies; NREL (National Renewable Energy Laboratory)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Melaina, Marc; Saur, Genevieve; Ramsden, Todd

2015-05-28

This presentation summarizes NREL's hydrogen and fuel cell analysis work in three areas: resource potential, greenhouse gas emissions and cost of delivered energy, and influence of auxiliary revenue streams. NREL's hydrogen and fuel cell analysis projects focus on low-­carbon and economic transportation and stationary fuel cell applications. Analysis tools developed by the lab provide insight into the degree to which bridging markets can strengthen the business case for fuel cell applications.

A custom-tailored FAMOS burn-up meter for VVER 440 fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simon, G.G.; Golochtchapov, S.; Glazov, A.G.

1995-12-31

The FAMOS fuel assembly monitoring system had been originally developed for monitoring irradiated fuel assemblies of the Karlsruhe Nuclear Research Center concentrating on neutron detection systems for special applications.The measurements in the past had demonstrated that FAMOS can perform precise measurements to control or measure with accuracy the main physical parameters of spent fuel. The FAMOS 3 system is specialized for burn-up determination of fuel assemblies. Thus it is possible to take into account the burn-up for the purposes of storage and transportation. The Kola NPP VVER 440 requirements necessitated developing an especially adopted FAMOS 3 system. In addition tomore » the passive neutron measurement, a gross gamma detection and a boron concentration monitoring system are implemented. The new system was constructed as well as tested in laboratory experiments. The monitoring system has been delivered to the customer and is ready for use.« less

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

Transfer of fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vuckovich, M.; Burkett, J. P.; Sallustio, J.

1984-12-11

Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If themore » strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint. This event occurs when a fuel assembly being deposited contacts the bottom of its slot or an obstruction in, or at the entry to the slot. To preclude lateral movement and possible damage to a fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge exceets the lov-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than the slack-line set-point. This occurs when the lines has been set in slack-line setting. When the line is tensioned after slack-li ne setting, the traverse drive remains enabled only if the line has been disconnected from the fuel assembly.« less

Polymer electrolyte membrane assembly for fuel cells

NASA Technical Reports Server (NTRS)

Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

2002-01-01

An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

Polymer electrolyte membrane assembly for fuel cells

NASA Technical Reports Server (NTRS)

Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

2000-01-01

An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

78 FR 9796 - Airworthiness Directives; Cessna Aircraft Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-12

... against the right steering tube assembly during rudder pedal actuation. This AD requires you to install... between the fuel return line assembly and the steering tube assembly and clearance between the fuel return...://www.cessnasupport.com . You may review copies of the referenced service information at the FAA, Small...

77 FR 50054 - Airworthiness Directives; Cessna Aircraft Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-20

... rubbing against the right steering tube assembly during full rudder pedal actuation. This proposed AD would require you to inspect the fuel return line assembly for chafing; replace the fuel return line... right steering tube assembly and the airplane structure; and adjustment as necessary. We are proposing...

77 FR 72250 - Airworthiness Directives; Cessna Aircraft Company Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-05

... rubbing against the right steering tube assembly during rudder pedal actuation. This proposed AD would require you to install the forward and aft fuel return line support clamps and brackets; inspect for a minimum clearance between the fuel return line assembly and the steering tube assembly and clearance...

CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

DOEpatents

Weems, S.J.

1963-09-24

>A rib arrangement and an end construction for nuclearfuel elements laid end to end in a coolant tube are described. The rib arrangement is such that each fuel element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube. The end construction of the fuel elements is such that the fuel elements when assembled end to end are keyed against relative rotation, and the ribs of each fuel element cooperate with the ribs of the adjacent fuel elements to give the assembled fuel elements a tight fit with the coolant tube. (AEC)

Alternative Fuels Data Center

Science.gov Websites

technologies and operational practices which increase fuel efficiency and reduce emissions from goods movement . EPA provides partners with performance benchmarking tools, fleet management best practices, technology is working with partners to test and verify advanced technologies and operational practices that save

Two-phase pressure drop reduction BWR assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.E.; Crowther, R.L.; Colby, M.J.

1991-05-21

This patent describes an improved fuel assembly for a boiling water reactor. It comprises: a fuel channel; a lower tie plate; an upper tie plate; the lower tie plate and the upper tie plate defining a two-dimensional matrix; at least one water rod the fuel rods being partial length rods.

75 FR 3876 - Mark Edward Leyse; Receipt of Petition for Rulemaking

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-25

... (assembly) severe fuel damage experiments. The petitioner also requests that the NRC promulgate a regulation... aware that data from multi-rod (assembly) severe fuel damage experiments indicates that the current... fuel damage experiments indicates that the current peak cladding temperature limit contained in 10 CFR...

Acceleration of boundary element method for linear elasticity

NASA Astrophysics Data System (ADS)

Zapletal, Jan; Merta, Michal; Čermák, Martin

2017-07-01

In this work we describe the accelerated assembly of system matrices for the boundary element method using the Intel Xeon Phi coprocessors. We present a model problem, provide a brief overview of its discretization and acceleration of the system matrices assembly using the coprocessors, and test the accelerated version using a numerical benchmark.

General view of a Space Shuttle Main Engine (SSME) mounted ...

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

General view of a Space Shuttle Main Engine (SSME) mounted on an SSME engine handler, taken in the SSME Processing Facility at Kennedy Space Center. The most prominent features of the engine assembly in this view are the Low-Pressure Fuel Turbopump Discharge Duct looping around the right side and underneath the assembly, the High-Pressure Fuel Turbopump located on the lower left portion of the assembly, the Engine Controller and Main Fuel Valve Hydraulic Actuator located on the upper portion of the assembly and the Low-Pressure Oxidizer Turbopump Discharge Duct at the top of the engine assembly in this view. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

PWR integral tie plate and locking mechanism

DOE Office of Scientific and Technical Information (OSTI.GOV)

Flora, B.S.; Osborne, J.L.

1980-08-26

A locking mechanism for securing an upper tie plate to the tie rods of a nuclear fuel bundle is described. The mechanism includes an upper tie plate assembly and locking sleeves fixed to the ends of the tie rods. The tie plate is part of the upper tie plate assembly and is secured to the fuel bundle by securing the entire upper tie plate assembly to the locking sleeves fixed to the tie rods. The assembly includes, in addition to the tie plate, locking nuts for engaging the locking sleeves, retaining sleeves to operably connect the locking nuts to themore » assembly, a spring biased reaction plate to restrain the locking nuts in the locked position and a means to facilitate the removal of the entire assembly as a unit from the fuel bundle.« less

Mixed Oxide Fresh Fuel Package Auxiliary Equipment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yapuncich, F.; Ross, A.; Clark, R.H.

2008-07-01

The United States Department of Energy's National Nuclear Security Administration (NNSA) is overseeing the construction the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) on the Savannah River Site. The new facility, being constructed by NNSA's contractor Shaw AREVA MOX Services, will fabricate fuel assemblies utilizing surplus plutonium as feedstock. The fuel will be used in designated commercial nuclear reactors. The MOX Fresh Fuel Package (MFFP), which has recently been licensed by the Nuclear Regulatory Commission (NRC) as a type B package (USA/9295/B(U)F-96), will be utilized to transport the fabricated fuel assemblies from the MFFF to the nuclear reactors. It wasmore » necessary to develop auxiliary equipment that would be able to efficiently handle the high precision fuel assemblies. Also, the physical constraints of the MFFF and the nuclear power plants require that the equipment be capable of loading and unloading the fuel assemblies both vertically and horizontally. The ability to reconfigure the load/unload evolution builds in a large degree of flexibility for the MFFP for the handling of many types of both fuel and non fuel payloads. The design and analysis met various technical specifications including dynamic and static seismic criteria. The fabrication was completed by three major fabrication facilities within the United States. The testing was conducted by Sandia National Laboratories. The unique design specifications and successful testing sequences will be discussed. (authors)« less

Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Serell, D.C.; Kaplan, S.

1980-09-01

Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.

Portable instrument for inspecting irradiated nuclear fuel assemblies

DOEpatents

Nicholson, Nicholas; Dowdy, Edward J.; Holt, David M.; Stump, Jr., Charles J.

1985-01-01

A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

Fuel Efficiency Mapping of a 2014 6-Cylinder GM EcoTec 4.3L Engine with Cylinder Deactivation (SAE 2016-01-0662)

EPA Science Inventory

This paper describes the method and test results of the engine dyno portion of the benchmarking test results including engine fuel consumption maps showing the effects of cylinder deactivation engine technology.

Method and apparatus for assembling solid oxide fuel cells

DOEpatents

Szreders, B.E.; Campanella, N.

1988-05-11

This invention relates generally to solid oxide fuel power generators and is particularly directed to improvements in the assembly and coupling of solid oxide fuel cell modules. A plurality of jet air tubes are supported and maintained in a spaced matrix array by a positioning/insertion assembly for insertion in respective tubes of a solid oxide fuel cell (SOFC) in the assembly of an SOFC module. The positioning/insertion assembly includes a plurality of generally planar, elongated, linear vanes which are pivotally mounted at each end thereof to a support frame. A rectangular compression assembly of adjustable size is adapted to receive and squeeze a matrix of SOFC tubes so as to compress the inter-tube nickel felt conductive pads which provide series/parallel electrical connection between adjacent SOFCs, with a series of increasingly larger retainer frames used to maintain larger matrices of SOFC tubes in position. Expansion of the SOFC module housing at the high operating temperatures of the SOFC is accommodated by conductive, flexible, resilient expansion, connector bars which provide support and electrical coupling at the top and bottom of the SOFC module housing. 17 figs.

PLATES WITH OXIDE INSERTS

DOEpatents

West, J.M.; Schumar, J.F.

1958-06-10

Planar-type fuel assemblies for nuclear reactors are described, particularly those comprising fuel in the oxide form such as thoria and urania. The fuel assembly consists of a plurality of parallel spaced fuel plate mennbers having their longitudinal side edges attached to two parallel supporting side plates, thereby providing coolant flow channels between the opposite faces of adjacent fuel plates. The fuel plates are comprised of a plurality of longitudinally extending tubular sections connected by web portions, the tubular sections being filled with a plurality of pellets of the fuel material and the pellets being thermally bonded to the inside of the tubular section by lead.

Implementation of Benchmarking Transportation Logistics Practices and Future Benchmarking Organizations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thrower, A.W.; Patric, J.; Keister, M.

2008-07-01

The purpose of the Office of Civilian Radioactive Waste Management's (OCRWM) Logistics Benchmarking Project is to identify established government and industry practices for the safe transportation of hazardous materials which can serve as a yardstick for design and operation of OCRWM's national transportation system for shipping spent nuclear fuel and high-level radioactive waste to the proposed repository at Yucca Mountain, Nevada. The project will present logistics and transportation practices and develop implementation recommendations for adaptation by the national transportation system. This paper will describe the process used to perform the initial benchmarking study, highlight interim findings, and explain how thesemore » findings are being implemented. It will also provide an overview of the next phase of benchmarking studies. The benchmarking effort will remain a high-priority activity throughout the planning and operational phases of the transportation system. The initial phase of the project focused on government transportation programs to identify those practices which are most clearly applicable to OCRWM. These Federal programs have decades of safe transportation experience, strive for excellence in operations, and implement effective stakeholder involvement, all of which parallel OCRWM's transportation mission and vision. The initial benchmarking project focused on four business processes that are critical to OCRWM's mission success, and can be incorporated into OCRWM planning and preparation in the near term. The processes examined were: transportation business model, contract management/out-sourcing, stakeholder relations, and contingency planning. More recently, OCRWM examined logistics operations of AREVA NC's Business Unit Logistics in France. The next phase of benchmarking will focus on integrated domestic and international commercial radioactive logistic operations. The prospective companies represent large scale shippers and have vast experience in safely and efficiently shipping spent nuclear fuel and other radioactive materials. Additional business processes may be examined in this phase. The findings of these benchmarking efforts will help determine the organizational structure and requirements of the national transportation system. (authors)« less

Hatch assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, J.R.; Hardin, R.T. Jr.

1987-07-07

This patent describes a nuclear reactor installation including means defining a fuel handling area and means defining a containment area separated from the fuel handling area and including a refuelling cavity; the improvement comprising: (a) a fuel transfer tube connecting the refuelling cavity with the fuel handling area; the fuel transfer tube having a first end in the fuel handling area and a second end in the refueling cavity; (b) valve means for opening and closing the first end; and (c) a hatch assembly mounted on the second end; the hatch assembly including (1) a hatch ring affixed to themore » fuel transfer tube at the second end the hatch ring has an integral annular seat surrounded by the hatch ring and defines a hatch opening in the second end of the fuel transfer tube; (2) a hatch cover adapts to be positioned on the annular seat for covering the hatch opening; (3) latching units are supported on the hatch ring about the hatch opening, each latching unit.« less

NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

DOEpatents

McCorkle, W.H.

1960-02-23

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

High efficiency gas burner

DOEpatents

Schuetz, Mark A.

1983-01-01

A burner assembly provides for 100% premixing of fuel and air by drawing the air into at least one high velocity stream of fuel without power assist. Specifically, the nozzle assembly for injecting the fuel into a throat comprises a plurality of nozzles in a generally circular array. Preferably, swirl is imparted to the air/fuel mixture by angling the nozzles. The diffuser comprises a conical primary diffuser followed by a cusp diffuser.

Detecting pin diversion from pressurized water reactors spent fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Young S.; Sitaraman, Shivakumar

Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and takingmore » the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.« less

Combustor with two stage primary fuel tube with concentric members and flow regulating

DOEpatents

Parker, David Marchant; Whidden, Graydon Lane; Zolyomi, Wendel

1999-01-01

A combustor for a gas turbine having a centrally located fuel nozzle and inner, middle and outer concentric cylindrical liners, the inner liner enclosing a primary combustion zone. The combustor has an air inlet that forms two passages for pre-mixing primary fuel and air to be supplied to the primary combustion zone. Each of the pre-mixing passages has a circumferential array of swirl vanes. A plurality of primary fuel tube assemblies extend through both pre-mixing passages, with each primary fuel tube assembly located between a pair of swirl vanes. Each primary fuel tube assembly is comprised of two tubular members. The first member supplies fuel to the first pre-mixing passage, while the second member, which extends through the first member, supplies fuel to the second pre-mixing passage. An annular fuel manifold is divided into first and second chambers by a circumferentially extending baffle. The proximal end of the first member is attached to the manifold itself while the proximal end of the second member is attached to the baffle. The distal end of the first member is attached directly to the second member at around its mid-point. The inlets of the first and second members are in flow communication with the first and second manifold chambers, respectively. Control valves separately regulate the flow of fuel to the two chambers and, therefore, to the two members of the fuel tube assemblies, thereby allowing the flow of fuel to the first and second pre-mixing passages to be separately controlled.

Evaluation of the Pool Critical Assembly Benchmark with Explicitly-Modeled Geometry using MCNP6

DOE PAGES

Kulesza, Joel A.; Martz, Roger Lee

2017-03-01

Despite being one of the most widely used benchmarks for qualifying light water reactor (LWR) radiation transport methods and data, no benchmark calculation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) pressure vessel wall benchmark facility (PVWBF) using MCNP6 with explicitly modeled core geometry exists. As such, this paper provides results for such an analysis. First, a criticality calculation is used to construct the fixed source term. Next, ADVANTG-generated variance reduction parameters are used within the final MCNP6 fixed source calculations. These calculations provide unadjusted dosimetry results using three sets of dosimetry reaction cross sections of varyingmore » ages (those packaged with MCNP6, from the IRDF-2002 multi-group library, and from the ACE-formatted IRDFF v1.05 library). These results are then compared to two different sets of measured reaction rates. The comparison agrees in an overall sense within 2% and on a specific reaction- and dosimetry location-basis within 5%. Except for the neptunium dosimetry, the individual foil raw calculation-to-experiment comparisons usually agree within 10% but is typically greater than unity. Finally, in the course of developing these calculations, geometry that has previously not been completely specified is provided herein for the convenience of future analysts.« less

Requirements to the procedure and stages of innovative fuel development

NASA Astrophysics Data System (ADS)

Troyanov, V.; Zabudko, L.; Grachyov, A.; Zhdanova, O.

2016-04-01

According to the accepted current understanding under the nuclear fuel we will consider the assembled active zone unit (Fuel assembly) with its structural elements, fuel rods, pellet column, structural materials of fuel rods and fuel assemblies. The licensing process includes justification of safe application of the proposed modifications, including design-basis and experimental justification of the modified items under normal operating conditions and in violation of normal conditions, including accidents as well. Besides the justification of modified units itself, it is required to show the influence of modifications on the performance and safety of the other Reactor Unit’ and Nuclear Plant’ elements (e.g. burst can detection system, transportation and processing operations during fuel handling), as well as to justify the new standards of fuel storage etc. Finally, the modified fuel should comply with the applicable regulations, which often becomes a very difficult task, if only because those regulations, such as the NP-082-07, are not covered modification issues. Making amendments into regulations can be considered as the only solution, but the process is complicated and requires deep grounds for amendments. Some aspects of licensing new nuclear fuel are considered the example of mixed nitride uranium -plutonium fuel application for the BREST reactor unit.

Storage, transportation and disposal system for used nuclear fuel assemblies

DOEpatents

Scaglione, John M.; Wagner, John C.

2017-01-10

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

Portable instrument for inspecting irradiated nuclear-fuel assemblies in a water-filled storage pond by measurement of induced Cerenkov radiation

DOEpatents

Nicholson, N.; Dowdy, E.J.; Holt, D.M.; Stump, C.J. Jr.

1982-05-13

A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

Benchmarking to improve the quality of cystic fibrosis care.

PubMed

Schechter, Michael S

2012-11-01

Benchmarking involves the ascertainment of healthcare programs with most favorable outcomes as a means to identify and spread effective strategies for delivery of care. The recent interest in the development of patient registries for patients with cystic fibrosis (CF) has been fueled in part by an interest in using them to facilitate benchmarking. This review summarizes reports of how benchmarking has been operationalized in attempts to improve CF care. Although certain goals of benchmarking can be accomplished with an exclusive focus on registry data analysis, benchmarking programs in Germany and the United States have supplemented these data analyses with exploratory interactions and discussions to better understand successful approaches to care and encourage their spread throughout the care network. Benchmarking allows the discovery and facilitates the spread of effective approaches to care. It provides a pragmatic alternative to traditional research methods such as randomized controlled trials, providing insights into methods that optimize delivery of care and allowing judgments about the relative effectiveness of different therapeutic approaches.

Mathematical model and metaheuristics for simultaneous balancing and sequencing of a robotic mixed-model assembly line

NASA Astrophysics Data System (ADS)

Li, Zixiang; Janardhanan, Mukund Nilakantan; Tang, Qiuhua; Nielsen, Peter

2018-05-01

This article presents the first method to simultaneously balance and sequence robotic mixed-model assembly lines (RMALB/S), which involves three sub-problems: task assignment, model sequencing and robot allocation. A new mixed-integer programming model is developed to minimize makespan and, using CPLEX solver, small-size problems are solved for optimality. Two metaheuristics, the restarted simulated annealing algorithm and co-evolutionary algorithm, are developed and improved to address this NP-hard problem. The restarted simulated annealing method replaces the current temperature with a new temperature to restart the search process. The co-evolutionary method uses a restart mechanism to generate a new population by modifying several vectors simultaneously. The proposed algorithms are tested on a set of benchmark problems and compared with five other high-performing metaheuristics. The proposed algorithms outperform their original editions and the benchmarked methods. The proposed algorithms are able to solve the balancing and sequencing problem of a robotic mixed-model assembly line effectively and efficiently.

Thermo-hydro-mechanical-chemical processes in fractured-porous media: Benchmarks and examples

NASA Astrophysics Data System (ADS)

Kolditz, O.; Shao, H.; Görke, U.; Kalbacher, T.; Bauer, S.; McDermott, C. I.; Wang, W.

2012-12-01

The book comprises an assembly of benchmarks and examples for porous media mechanics collected over the last twenty years. Analysis of thermo-hydro-mechanical-chemical (THMC) processes is essential to many applications in environmental engineering, such as geological waste deposition, geothermal energy utilisation, carbon capture and storage, water resources management, hydrology, even climate change. In order to assess the feasibility as well as the safety of geotechnical applications, process-based modelling is the only tool to put numbers, i.e. to quantify future scenarios. This charges a huge responsibility concerning the reliability of computational tools. Benchmarking is an appropriate methodology to verify the quality of modelling tools based on best practices. Moreover, benchmarking and code comparison foster community efforts. The benchmark book is part of the OpenGeoSys initiative - an open source project to share knowledge and experience in environmental analysis and scientific computation.

Advanced Fuel Cell System Thermal Management for NASA Exploration Missions

NASA Technical Reports Server (NTRS)

Burke, Kenneth A.

2009-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA exploration program. An analysis of a state-of-the-art fuel cell cooling systems was done to benchmark the portion of a fuel cell system s mass that is dedicated to thermal management. Additional analysis was done to determine the key performance targets of the advanced passive thermal management technology that would substantially reduce fuel cell system mass.

Katome: de novo DNA assembler implemented in rust

NASA Astrophysics Data System (ADS)

Neumann, Łukasz; Nowak, Robert M.; Kuśmirek, Wiktor

2017-08-01

Katome is a new de novo sequence assembler written in the Rust programming language, designed with respect to future parallelization of the algorithms, run time and memory usage optimization. The application uses new algorithms for the correct assembly of repetitive sequences. Performance and quality tests were performed on various data, comparing the new application to `dnaasm', `ABySS' and `Velvet' genome assemblers. Quality tests indicate that the new assembler creates more contigs than well-established solutions, but the contigs have better quality with regard to mismatches per 100kbp and indels per 100kbp. Additionally, benchmarks indicate that the Rust-based implementation outperforms `dnaasm', `ABySS' and `Velvet' assemblers, written in C++, in terms of assembly time. Lower memory usage in comparison to `dnaasm' is observed.

Fuel cell with electrolyte matrix assembly

DOEpatents

Kaufman, Arthur; Pudick, Sheldon; Wang, Chiu L.

1988-01-01

This invention is directed to a fuel cell employing a substantially immobilized electrolyte imbedded therein and having a laminated matrix assembly disposed between the electrodes of the cell for holding and distributing the electrolyte. The matrix assembly comprises a non-conducting fibrous material such as silicon carbide whiskers having a relatively large void-fraction and a layer of material having a relatively small void-fraction.

Modular fuel-cell stack assembly

DOEpatents

Patel, Pinakin [Danbury, CT; Urko, Willam [West Granby, CT

2008-01-29

A modular multi-stack fuel-cell assembly in which the fuel-cell stacks are situated within a containment structure and in which a gas distributor is provided in the structure and distributes received fuel and oxidant gases to the stacks and receives exhausted fuel and oxidant gas from the stacks so as to realize a desired gas flow distribution and gas pressure differential through the stacks. The gas distributor is centrally and symmetrically arranged relative to the stacks so that it itself promotes realization of the desired gas flow distribution and pressure differential.

Precise calculation of neutron-capture reactions contribution in energy release for different types of VVER-1000 fuel assemblies

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Bahdanovich, Rynat; Pham, Phu

2017-09-01

Precise calculation of energy release in a nuclear reactor is necessary to obtain the correct spatial power distribution and predict characteristics of burned nuclear fuel. In this work, previously developed method for calculation neutron-capture reactions - capture component - contribution in effective energy release in a fuel core of nuclear reactor is discussed. The method was improved and implemented to the different models of VVER-1000 reactor developed for MCU 5 and MCNP 4 computer codes. Different models of equivalent cell and fuel assembly in the beginning of fuel cycle were calculated. These models differ by the geometry, fuel enrichment and presence of burnable absorbers. It is shown, that capture component depends on fuel enrichment and presence of burnable absorbers. Its value varies for different types of hot fuel assemblies from 3.35% to 3.85% of effective energy release. Average capture component contribution in effective energy release for typical serial fresh fuel of VVER-1000 is 3.5%, which is 7 MeV/fission. The method will be used in future to estimate the dependency of capture energy on fuel density, burn-up, etc.

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE PAGES

Ilas, Germina; Liljenfeldt, Henrik

2017-05-19

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Liljenfeldt, Henrik

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

General view of the Space Shuttle Main Engine (SSME) assembly ...

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

General view of the Space Shuttle Main Engine (SSME) assembly with the expansion nozzle removed and resting on a cushioned mat on the floor of the SSME Processing Facility. The most prominent features in this view are the Low-pressure oxidizer Turbopump discharge Duct looping from the upper left side of the engine assembly to the lower left side of the assembly, the Low-Pressure Fuel Turbopump (LPFTP) is on the upper left of the assembly in this view and the LPFTP Discharge Duct loops from the upper left to upper right then turns back and down the assembly to the High-Pressure Fuel Turbopump on the lower right of the assembly. The Engine Controller and the Main fuel Valve Hydraulic Actuator are on the lower left portion of the assembly. The vertical rod that is in the approximate center of the engine assembly is a piece of ground support equipment call a Gimbal Actuator Replacement Strut which are used on the SSMEs when they are not installed in an orbiter. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

PBF Reactor Building (PER620). Fuel rod test assembly is on ...

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

PBF Reactor Building (PER-620). Fuel rod test assembly is on display at PBF. Date: 1982. INEEL negative no. 82-4893 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Membrane electrode assembly for a fuel cell

NASA Technical Reports Server (NTRS)

Prakash, Surya (Inventor); Narayanan, Sekharipuram R. (Inventor); Atti, Anthony (Inventor); Olah, George (Inventor); Smart, Marshall C. (Inventor)

2006-01-01

A catalyst ink for a fuel cell including a catalytic material and poly(vinylidene fluoride). The ink may be applied to a substrate to form an electrode, or bonded with other electrode layers to form a membrane electrode assembly (MEA).

Method of bonding

DOEpatents

Saller, deceased, Henry A.; Hodge, Edwin S.; Paprocki, Stanley J.; Dayton, Russell W.

1987-12-01

1. A method of making a fuel-containing structure for nuclear reactors, comprising providing an assembly comprising a plurality of fuel units; each fuel unit consisting of a core plate containing thermal-neutron-fissionable material, sheets of cladding metal on its bottom and top surfaces, said cladding sheets being of greater width and length than said core plates whereby recesses are formed at the ends and sides of said core plate, and end pieces and first side pieces of cladding metal of the same thickness as the core plate positioned in said recesses, the assembly further comprising a plurality of second side pieces of cladding metal engaging the cladding sheets so as to space the fuel units from one another, and a plurality of filler plates of an acid-dissolvable nonresilient material whose melting point is above 2000.degree. F., each filler plate being arranged between a pair of said second side pieces and the cladding plates of two adjacent fuel units, the filler plates having the same thickness as the second side pieces; the method further comprising enclosing the entire assembly in an envelope; evacuating the interior of the entire assembly through said envelope; applying inert gas under a pressure of about 10,000 psi to the outside of said envelope while at the same time heating the assembly to a temperature above the flow point of the cladding metal but below the melting point of any material of the assembly, whereby the envelope is pressed against the assembly and integral bonds are formed between plates, sheets, first side pieces, and end pieces and between the sheets and the second side pieces; slowly cooling the assembly to room temperature; removing the envelope; and dissolving the filler plates without attacking the cladding metal.

Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

DOE PAGES

Marshall, Margaret A.

2014-11-04

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with themore » GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.« less

Storage, transportation and disposal system for used nuclear fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scaglione, John M.; Wagner, John C.

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. Themore » system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.« less

Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

NASA Astrophysics Data System (ADS)

Anghaie, Samim; Knight, Travis; Gouw, Reza; Furman, Eric

2001-02-01

Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a grooved graphite tube to form a SLHC fuel assembly. A total of 18 fuel assemblies are arranged circumferentially to form two concentric rings of fuel assemblies with zirconium hydride filling the space between assemblies. For 50 to 250 kilo Newtons thrust operations, the reactor diameter and length including reflectors are 57 cm and 60 cm, respectively. Results of the nuclear design and thermal fluid analyses of the SLHC nuclear thermal propulsion system are presented. .

In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauld, Ian C.; Hu, Jianwei; De Baere, P.

Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the frameworkmore » of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel verification technique.« less

Fuel burner and combustor assembly for a gas turbine engine

DOEpatents

Leto, Anthony

1983-01-01

A fuel burner and combustor assembly for a gas turbine engine has a housing within the casing of the gas turbine engine which housing defines a combustion chamber and at least one fuel burner secured to one end of the housing and extending into the combustion chamber. The other end of the fuel burner is arranged to slidably engage a fuel inlet connector extending radially inwardly from the engine casing so that fuel is supplied, from a source thereof, to the fuel burner. The fuel inlet connector and fuel burner coact to anchor the housing against axial movement relative to the engine casing while allowing relative radial movement between the engine casing and the fuel burner and, at the same time, providing fuel flow to the fuel burner. For dual fuel capability, a fuel injector is provided in said fuel burner with a flexible fuel supply pipe so that the fuel injector and fuel burner form a unitary structure which moves with the fuel burner.

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Favalli, Andrea; Vo, D.; Grogan, Brandon R.

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

DOE PAGES

Favalli, Andrea; Vo, D.; Grogan, Brandon R.; ...

2016-02-26

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Temporal switching of an amphiphilic self-assembly by a chemical fuel-driven conformational response† †Electronic supplementary information (ESI) available: Experimental section, synthetic procedures and supporting figures. See DOI: 10.1039/c7sc01730h Click here for additional data file.

PubMed Central

Jalani, Krishnendu; Dhiman, Shikha

2017-01-01

The spatial and temporal control of self-assemblies is the latest scientific hurdle in supramolecular chemistry which is inspired by the functioning of biological systems fueled by chemical signals. In this study, we work towards alleviating this scenario by employing a unique amphiphilic foldamer that operates under the effect of a chemical fuel. The conformational changes in the foldamer amplify into observable morphological changes in its amphiphilic assembly that are controlled by external molecular cues (fuel). We take advantage of this redox responsive foldamer to affect its conformation in a temporal manner by an enzymatic pathway. The temporal characteristics of the transient conformation/assembly can be modulated by varying the concentrations of the fuel and enzyme. We believe that such a design strategy can have positive consequences in designing molecular and supramolecular systems for future active, adaptive and autonomous materials. PMID:28989632

Direct methanol feed fuel cell with reduced catalyst loading

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor)

1999-01-01

Improvements to direct feed methanol fuel cells include new protocols for component formation. Catalyst-water repellent material is applied in formation of electrodes and sintered before application of ionomer. A membrane used in formation of an electrode assembly is specially pre-treated to improve bonding between catalyst and membrane. The improved electrode and the pre-treated membrane are assembled into a membrane electrode assembly.

Co-flow anode/cathode supply heat exchanger for a solid-oxide fuel cell assembly

DOEpatents

Haltiner, Jr., Karl J.; Kelly, Sean M.

2005-11-22

In a solid-oxide fuel cell assembly, a co-flow heat exchanger is provided in the flow paths of the reformate gas and the cathode air ahead of the fuel cell stack, the reformate gas being on one side of the exchanger and the cathode air being on the other. The reformate gas is at a substantially higher temperature than is desired in the stack, and the cathode gas is substantially cooler than desired. In the co-flow heat exchanger, the temperatures of the reformate and cathode streams converge to nearly the same temperature at the outlet of the exchanger. Preferably, the heat exchanger is formed within an integrated component manifold (ICM) for a solid-oxide fuel cell assembly.

Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope; R. Sonat Sen; Brian Boer

2011-09-01

The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code tomore » assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilkey, Lindsay

This milestone presents a demonstration of the High-to-Low (Hi2Lo) process in the VVI focus area. Validation and additional calculations with the commercial computational fluid dynamics code, STAR-CCM+, were performed using a 5x5 fuel assembly with non-mixing geometry and spacer grids. This geometry was based on the benchmark experiment provided by Westinghouse. Results from the simulations were compared to existing experimental data and to the subchannel thermal-hydraulics code COBRA-TF (CTF). An uncertainty quantification (UQ) process was developed for the STAR-CCM+ model and results of the STAR UQ were communicated to CTF. Results from STAR-CCM+ simulations were used as experimental design pointsmore » in CTF to calibrate the mixing parameter β and compared to results obtained using experimental data points. This demonstrated that CTF’s β parameter can be calibrated to match existing experimental data more closely. The Hi2Lo process for the STAR-CCM+/CTF code coupling was documented in this milestone and closely linked L3:VVI.H2LP15.01 milestone report.« less

Proton Exchange Membrane Fuel Cell Engineering Model Powerplant. Test Report: Benchmark Tests in Three Spatial Orientations

NASA Technical Reports Server (NTRS)

Loyselle, Patricia; Prokopius, Kevin

2011-01-01

Proton exchange membrane (PEM) fuel cell technology is the leading candidate to replace the aging alkaline fuel cell technology, currently used on the Shuttle, for future space missions. This test effort marks the final phase of a 5-yr development program that began under the Second Generation Reusable Launch Vehicle (RLV) Program, transitioned into the Next Generation Launch Technologies (NGLT) Program, and continued under Constellation Systems in the Exploration Technology Development Program. Initially, the engineering model (EM) powerplant was evaluated with respect to its performance as compared to acceptance tests carried out at the manufacturer. This was to determine the sensitivity of the powerplant performance to changes in test environment. In addition, a series of tests were performed with the powerplant in the original standard orientation. This report details the continuing EM benchmark test results in three spatial orientations as well as extended duration testing in the mission profile test. The results from these tests verify the applicability of PEM fuel cells for future NASA missions. The specifics of these different tests are described in the following sections.

Summary of BISON Development and Validation Activities - NEAMS FY16 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, R. L.; Pastore, G.; Gamble, K. A.

This summary report contains an overview of work performed under the work package en- titled “FY2016 NEAMS INL-Engineering Scale Fuel Performance (BISON)” A first chapter identifies the specific FY-16 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments, which in- clude: 1) publication of a journal article on solution verification and validation of BISON for LWR fuel, 2) publication of a journal article on 3D Missing Pellet Surface (MPS) analysis of BWR fuel, 3) use of BISON to designmore » a unique 3D MPS validation experiment for future in- stallation in the Halden research reactor, 4) participation in an OECD benchmark on Pellet Clad Mechanical Interaction (PCMI), 5) participation in an OECD benchmark on Reactivity Insertion Accident (RIA) analysis, 6) participation in an OECD activity on uncertainity quantification and sensitivity analysis in nuclear fuel modeling and 7) major improvements to BISON’s fission gas behavior models. A final chapter outlines FY-17 future work.« less

Compliant fuel cell system

DOEpatents

Bourgeois, Richard Scott [Albany, NY; Gudlavalleti, Sauri [Albany, NY

2009-12-15

A fuel cell assembly comprising at least one metallic component, at least one ceramic component and a structure disposed between the metallic component and the ceramic component. The structure is configured to have a lower stiffness compared to at least one of the metallic component and the ceramic component, to accommodate a difference in strain between the metallic component and the ceramic component of the fuel cell assembly.

Solution of the neutronics code dynamic benchmark by finite element method

NASA Astrophysics Data System (ADS)

Avvakumov, A. V.; Vabishchevich, P. N.; Vasilev, A. O.; Strizhov, V. F.

2016-10-01

The objective is to analyze the dynamic benchmark developed by Atomic Energy Research for the verification of best-estimate neutronics codes. The benchmark scenario includes asymmetrical ejection of a control rod in a water-type hexagonal reactor at hot zero power. A simple Doppler feedback mechanism assuming adiabatic fuel temperature heating is proposed. The finite element method on triangular calculation grids is used to solve the three-dimensional neutron kinetics problem. The software has been developed using the engineering and scientific calculation library FEniCS. The matrix spectral problem is solved using the scalable and flexible toolkit SLEPc. The solution accuracy of the dynamic benchmark is analyzed by condensing calculation grid and varying degree of finite elements.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Sterbentz, James W.; Snoj, Luka

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

PBF Reactor Building (PER620). PBF crane holds fuel test assembly ...

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

PBF Reactor Building (PER-620). PBF crane holds fuel test assembly aloft prior to lowering into reactor for test. Date: 1982. INEEL negative no. 82-4909 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

System for fuel rod removal from a reactor module

DOEpatents

Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

1988-07-28

A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

System for fuel rod removal from a reactor module

DOEpatents

Matchett, Richard L.; Roof, David R.; Kikta, Thomas J.; Wilczynski, Rosemarie; Nilsen, Roy J.; Bacvinskas, William S.; Fodor, George

1990-01-01

A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

Method for monitoring irradiated fuel using Cerenkov radiation

DOEpatents

Dowdy, E.J.; Nicholson, N.; Caldwell, J.T.

1980-05-21

A method is provided for monitoring irradiated nuclear fuel inventories located in a water-filled storage pond wherein the intensity of the Cerenkov radiation emitted from the water in the vicinity of the nuclear fuel is measured. This intensity is then compared with the expected intensity for nuclear fuel having a corresponding degree of irradiation exposure and time period after removal from a reactor core. Where the nuclear fuel inventory is located in an assembly having fuel pins or rods with intervening voids, the Cerenkov light intensity measurement is taken at selected bright sports corresponding to the water-filled interstices of the assembly in the water storage, the water-filled interstices acting as Cerenkov light channels so as to reduce cross-talk. On-line digital analysis of an analog video signal is possible, or video tapes may be used for later measurement using a video editor and an electrometer. Direct measurement of the Cerenkov radiation intensity also is possible using spot photometers pointed at the assembly.

Pulverized fuel-oxygen burner

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor, Curtis; Patterson, Brad; Perdue, Jayson

A burner assembly combines oxygen and fuel to produce a flame. The burner assembly includes an oxygen supply tube adapted to receive a stream of oxygen and a solid fuel conduit arranged to extend through the oxygen tube to convey a stream of fluidized, pulverized, solid fuel into a flame chamber. Oxygen flowing through the oxygen supply tube passes generally tangentially through a first set of oxygen-injection holes formed in the solid fuel conduit and off-tangentially from a second set of oxygen-injection holes formed in the solid fuel conduit and then mixes with fluidized, pulverized, solid fuel passing through themore » solid fuel conduit to create an oxygen-fuel mixture in a downstream portion of the solid fuel conduit. This mixture is discharged into a flame chamber and ignited in the flame chamber to produce a flame.« less

A comprehensive evaluation of different radiation models in a gas turbine combustor under conditions of oxy-fuel combustion with dry recycle

NASA Astrophysics Data System (ADS)

Kez, V.; Liu, F.; Consalvi, J. L.; Ströhle, J.; Epple, B.

2016-03-01

The oxy-fuel combustion is a promising CO2 capture technology from combustion systems. This process is characterized by much higher CO2 concentrations in the combustion system compared to that of the conventional air-fuel combustion. To accurately predict the enhanced thermal radiation in oxy-fuel combustion, it is essential to take into account the non-gray nature of gas radiation. In this study, radiation heat transfer in a 3D model gas turbine combustor under two test cases at 20 atm total pressure was calculated by various non-gray gas radiation models, including the statistical narrow-band (SNB) model, the statistical narrow-band correlated-k (SNBCK) model, the wide-band correlated-k (WBCK) model, the full spectrum correlated-k (FSCK) model, and several weighted sum of gray gases (WSGG) models. Calculations of SNB, SNBCK, and FSCK were conducted using the updated EM2C SNB model parameters. Results of the SNB model are considered as the benchmark solution to evaluate the accuracy of the other models considered. Results of SNBCK and FSCK are in good agreement with the benchmark solution. The WBCK model is less accurate than SNBCK or FSCK. Considering the three formulations of the WBCK model, the multiple gases formulation is the best choice regarding the accuracy and computational cost. The WSGG model with the parameters of Bordbar et al. (2014) [20] is the most accurate of the three investigated WSGG models. Use of the gray WSSG formulation leads to significant deviations from the benchmark data and should not be applied to predict radiation heat transfer in oxy-fuel combustion systems. A best practice to incorporate the state-of-the-art gas radiation models for high accuracy of radiation heat transfer calculations at minimal increase in computational cost in CFD simulation of oxy-fuel combustion systems for pressure path lengths up to about 10 bar m is suggested.

A NEUTRONIC REACTOR

DOEpatents

Luebke, E.A.; Vandenberg, L.B.

1959-09-01

A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1979-10-01

The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminary design for a high-burnup fuel rod, physics analyses of extended-burnup fuel cycles, studies of the physics characteristics of changes in fuel assembly metal-to-water ratios, and developmentmore » of a design concept for post-irradiation examination equipment to be utilized in examining high-burnup lead-test assemblies.« less

Present Status and Extensions of the Monte Carlo Performance Benchmark

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Petrovic, Bojan; Martin, William R.

2014-06-01

The NEA Monte Carlo Performance benchmark started in 2011 aiming to monitor over the years the abilities to perform a full-size Monte Carlo reactor core calculation with a detailed power production for each fuel pin with axial distribution. This paper gives an overview of the contributed results thus far. It shows that reaching a statistical accuracy of 1 % for most of the small fuel zones requires about 100 billion neutron histories. The efficiency of parallel execution of Monte Carlo codes on a large number of processor cores shows clear limitations for computer clusters with common type computer nodes. However, using true supercomputers the speedup of parallel calculations is increasing up to large numbers of processor cores. More experience is needed from calculations on true supercomputers using large numbers of processors in order to predict if the requested calculations can be done in a short time. As the specifications of the reactor geometry for this benchmark test are well suited for further investigations of full-core Monte Carlo calculations and a need is felt for testing other issues than its computational performance, proposals are presented for extending the benchmark to a suite of benchmark problems for evaluating fission source convergence for a system with a high dominance ratio, for coupling with thermal-hydraulics calculations to evaluate the use of different temperatures and coolant densities and to study the correctness and effectiveness of burnup calculations. Moreover, other contemporary proposals for a full-core calculation with realistic geometry and material composition will be discussed.

Two-phase pressure drop reduction BWR assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.E.; Crowther, R.L.; Colby, M.J.

1992-05-12

This patent describes a boiling water reactor having discrete bundles of fuel rods confined within channel enclosed fuel assemblies, an improvement to a fuel bundle assembly for placement in the reactor. It comprises a fuel channel having vertically extending walls forming a continuous channel around a fuel assembly volume, the channel being open at the bottom end for engagement to a lower tie plate and open at the upper end for engagement to an upper tie plate; rods for placement within the chamber, each the rod containing fissile material for producing nuclear reaction when in the presence of sufficient moderatedmore » neutron flux; a lower tie plate for supporting the bundle of rods within the channel, the lower tie plate for supporting the bundle of rods within the channel, the lower tie plate joining the bottom of the channel to close the bottom end of the channel, the lower tie plate providing defined apertures for the inflow of water in the channel between the rods for the generating of steam during the nuclear reaction; the plurality of fuel rods extending from the lower tie plate wherein a single phase region of the water in the bundle is defined to an upward portion of the bundle wherein a two phase region of the water and steam in the bundle is defined during nuclear steam generating reaction in the fuel bundle.« less

A photocatalyst-enzyme coupled artificial photosynthesis system for solar energy in production of formic acid from CO2.

PubMed

Yadav, Rajesh K; Baeg, Jin-Ook; Oh, Gyu Hwan; Park, No-Joong; Kong, Ki-jeong; Kim, Jinheung; Hwang, Dong Won; Biswas, Soumya K

2012-07-18

The photocatalyst-enzyme coupled system for artificial photosynthesis process is one of the most promising methods of solar energy conversion for the synthesis of organic chemicals or fuel. Here we report the synthesis of a novel graphene-based visible light active photocatalyst which covalently bonded the chromophore, such as multianthraquinone substituted porphyrin with the chemically converted graphene as a photocatalyst of the artificial photosynthesis system for an efficient photosynthetic production of formic acid from CO(2). The results not only show a benchmark example of the graphene-based material used as a photocatalyst in general artificial photosynthesis but also the benchmark example of the selective production system of solar chemicals/solar fuel directly from CO(2).

Air Breathing Direct Methanol Fuel Cell

DOEpatents

Ren; Xiaoming

2003-07-22

A method for activating a membrane electrode assembly for a direct methanol fuel cell is disclosed. The method comprises operating the fuel cell with humidified hydrogen as the fuel followed by running the fuel cell with methanol as the fuel.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.

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 investigationmore » 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 15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.« less

Advancing the Fork detector for quantitative spent nuclear fuel verification

DOE PAGES

Vaccaro, S.; Gauld, I. C.; Hu, J.; ...

2018-01-31

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. Finally, the results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.« less

Advancing the Fork detector for quantitative spent nuclear fuel verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Gauld, I. C.; Hu, J.

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. Finally, the results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.« less

Advancing the Fork detector for quantitative spent nuclear fuel verification

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Gauld, I. C.; Hu, J.; De Baere, P.; Peterson, J.; Schwalbach, P.; Smejkal, A.; Tomanin, A.; Sjöland, A.; Tobin, S.; Wiarda, D.

2018-04-01

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations. A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This paper describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. The results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.

MoMaS reactive transport benchmark using PFLOTRAN

NASA Astrophysics Data System (ADS)

Park, H.

2017-12-01

MoMaS benchmark was developed to enhance numerical simulation capability for reactive transport modeling in porous media. The benchmark was published in late September of 2009; it is not taken from a real chemical system, but realistic and numerically challenging tests. PFLOTRAN is a state-of-art massively parallel subsurface flow and reactive transport code that is being used in multiple nuclear waste repository projects at Sandia National Laboratories including Waste Isolation Pilot Plant and Used Fuel Disposition. MoMaS benchmark has three independent tests with easy, medium, and hard chemical complexity. This paper demonstrates how PFLOTRAN is applied to this benchmark exercise and shows results of the easy benchmark test case which includes mixing of aqueous components and surface complexation. Surface complexations consist of monodentate and bidentate reactions which introduces difficulty in defining selectivity coefficient if the reaction applies to a bulk reference volume. The selectivity coefficient becomes porosity dependent for bidentate reaction in heterogeneous porous media. The benchmark is solved by PFLOTRAN with minimal modification to address the issue and unit conversions were made properly to suit PFLOTRAN.

Method and apparatus for selectively controlling the speed of an engine

DOEpatents

Davis, Roy Inge

2001-02-27

A control assembly 12 for use within a vehicle 10 having an engine 14 and which selectively controls the speed of the engine 14 in order to increase fuel efficiency and to effect relatively smooth starting and stopping of the engine. Particularly, in one embodiment, control assembly 12 cooperatively operates with a starter/alternator assembly 20 and is adapted for use with hybrid vehicles employing a start/stop powertrain assembly, wherein fuel efficiency is increased by selectively stopping engine operation when the vehicle has stopped.

Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahman, Fariz Abdul; Lee, John C.; Franceschini, Fausto

2012-07-01

As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning themore » legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and burn TRU in a thermal spectrum, while satisfying top-level operational and safety constraints. Various assembly designs have been proposed to assess the TRU burning potential of Th-based fuel in PWRs. In addition to typical homogeneous loading patterns, heterogeneous configurations exploiting the breeding potential of thorium to enable multiple cycles of TRU irradiation and burning have been devised. The homogeneous assembly design, with all pins featuring TRU in Th, has the benefit of a simple loading pattern and the highest rate of TRU transmutation, but it can be used only for a few cycles due to the rapid rise in the TRU content of the recycled fuel, which challenges reactivity control, safety coefficients and fuel handling. Due to its simple loading pattern, such assembly design can be used as the first step of Th implementation, achieving up to 3 times larger TRU transmutation rate than conventional U-MOX, assuming same fraction of MOX assemblies in the core. As the next step in thorium implementation, heterogeneous assemblies featuring a mixed array of Th-U and Th-U-TRU pins, where the U is in-bred from Th, have been proposed. These designs have the potential to enable burning an external supply of TRU through multiple cycles of irradiation, recovery (via reprocessing) and recycling of the residual actinides at the end of each irradiation cycle. This is achieved thanks to a larger breeding of U from Th in the heterogeneous assemblies, which reduces the TRU supply and thus mitigates the increase in the TRU core inventory for the multi-recycled fuel. While on an individual cycle basis the amount of TRU burned in the heterogeneous assembly is reduced with respect to the homogeneous design, TRU burning rates higher than single-pass U-MOX fuel can still be achieved, with the additional benefits of a multi-cycle transmutation campaign recycling all TRU isotopes. Nitride fuel, due its higher density and U breeding potential, together with its better thermal properties, ideally suits the objectives and constraints of the heterogeneous assemblies. However, significant technological advancements must be made before nitride fuels can be employed in an LWR: its water resistance needs to be improved and a viable technology to enrich N in N-15 must be devised. Moreover, for the nitride heterogeneous configurations examined in this study, the enhancement in TRU burning performance is achieved not only by replacing oxide with nitride fuel, but also by increasing the fuel rod size. This latter modification, allowed by the high thermal conductivity of nitride fuel, leads however to a very tight lattice, which may challenge reactor coolant pumps and assembly hold-down mechanisms, the former through an increase in core pressure drop and the latter through an increase in assembly lift-off forces. To alleviate these issues, while still achieving the large fuel-to-moderator ratios resulting from using tight lattices, wire wraps could be used in place of grid spacers. For tight lattices, typical grid spacers are hard to manufacture and their replacement with wire wraps is known to allow for a pressure drop reduction by at least 2 times. The studies, while certainly very preliminary, provide a starting point to devise an optimum strategy for TRU transmutation in Th-based PWR fuel. The viability of the scheme proposed depends on the timely phasing in of the associated technologies, with proper lead time and to solve the many challenges. These challenges are certainly substantial, and make the current once-through U-based scheme pursued in the US by far a more practical (and cheaper) option. However, when compared to other transmutation schemes, the proposed one has arguably similar challenges and unknowns with potentially bigger rewards. (authors)« less

CHF considerations for highly moderated 100% MOX fuels PWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saphier, D.; Raymond, P.

1995-09-01

A feasibility study on using 100% MOX fuel in a PWR with increased moderating ratio, RMA, was initiated. In the proposed design all the parameters were chosen identical to the French 1450MW PWR, except the fuel pin diameter which was reduced to achieve higher moderating ratios, V{sub M}/V{sub F}, where V{sub M} and V{sub F} are the moderator and fuel volume respectively. Moderating ratios from 2 to 4 were considered. In the present study the thermal-hydraulic feasibility of using fuel assemblies with smaller diameter fuel pins was investigated. The major design constrain in this study was the critical heat fluxmore » (CHF). In order to maintain the fuel pin integrity under nominal operating and transient conditions, the minimum DNBR, (Departure from Nucleate Boiling Ratio given by CHF/q{close_quotes}{sub local}, where q{close_quotes}{sub local} is the local heat flux), has to be above a given value. The limitations of the existing CHF correlations for the present study are outlined. Two designs based on the conventional 17x17 fuel assembly and on the advanced 19x19 assembly meeting the MDNBR criteria and satisfying the control margin requirements, are proposed.« less

Fuel-air munition and device

DOEpatents

Carlson, Gary A.

1976-01-01

An aerially delivered fuel-air munition consisting of an impermeable tank filled with a pressurized liquid fuel and joined at its two opposite ends with a nose section and a tail assembly respectively to complete an aerodynamic shape. On impact the tank is explosively ruptured to permit dispersal of the fuel in the form of a fuel-air cloud which is detonated after a preselected time delay by means of high explosive initiators ejected from the tail assembly. The primary component in the fuel is methylacetylene, propadiene, or mixtures thereof to which is added a small mole fraction of a relatively high vapor pressure liquid diluent or a dissolved gas diluent having a low solubility in the primary component.

Fuel cell with electrolyte feed system

DOEpatents

Feigenbaum, Haim

1984-01-01

A fuel cell having a pair of electrodes at the sites of electrochemical reactions of hydrogen and oxygen and a phosphoric acid electrolyte provided with an electrolyte supporting structure in the form of a laminated matrix assembly disposed between the electrodes. The matrix assembly is formed of a central layer disposed between two outer layers, each being permeable to the flow of the electrolyte. The central layer is provided with relatively large pores while the outer layers are provided with relatively small pores. An external reservoir supplies electrolyte via a feed means to the central layer to compensate for changes in electrolyte volume in the matrix assembly during the operation of fuel cell.

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

DOE PAGES

Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.; ...

2017-10-07

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

NASA Astrophysics Data System (ADS)

Menlove, Howard; Belian, Anthony; Geist, William; Rael, Carlos

2018-01-01

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd2O3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate the 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. We have named the system the fast-neutron passive collar (FNPC).

Fuel cell system with separating structure bonded to electrolyte

DOEpatents

Bourgeois, Richard Scott; Gudlavalleti, Sauri; Quek, Shu Ching; Hasz, Wayne Charles; Powers, James Daniel

2010-09-28

A fuel cell assembly comprises a separating structure configured for separating a first reactant and a second reactant wherein the separating structure has an opening therein. The fuel cell assembly further comprises a fuel cell comprising a first electrode, a second electrode, and an electrolyte interposed between the first and second electrodes, and a passage configured to introduce the second reactant to the second electrode. The electrolyte is bonded to the separating structure with the first electrode being situated within the opening, and the second electrode being situated within the passage.

NEAMS update quarterly report for January - March 2012.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bradley, K.S.; Hayes, S.; Pointer, D.

Quarterly highlights are: (1) The integration of Denovo and AMP was demonstrated in an AMP simulation of the thermo-mechanics of a complete fuel assembly; (2) Bison was enhanced with a mechanistic fuel cracking model; (3) Mechanistic algorithms were incorporated into various lower-length-scale models to represent fission gases and dislocations in UO2 fuels; (4) Marmot was improved to allow faster testing of mesoscale models using larger problem domains; (5) Component models of reactor piping were developed for use in Relap-7; (6) The mesh generator of Proteus was updated to accept a mesh specification from Moose and equations were formulated for themore » intermediate-fidelity Proteus-2D1D module; (7) A new pressure solver was implemented in Nek5000 and demonstrated to work 2.5 times faster than the previous solver; (8) Work continued on volume-holdup models for two fuel reprocessing operations: voloxidation and dissolution; (9) Progress was made on a pyroprocessing model and the characterization of pyroprocessing emission signatures; (10) A new 1D groundwater waste transport code was delivered to the used fuel disposition (UFD) campaign; (11) Efforts on waste form modeling included empirical simulation of sodium-borosilicate glass compositions; (12) The Waste team developed three prototypes for modeling hydride reorientation in fuel cladding during very long-term fuel storage; (13) A benchmark demonstration problem (fission gas bubble growth) was modeled to evaluate the capabilities of different meso-scale numerical methods; (14) Work continued on a hierarchical up-scaling framework to model structural materials by directly coupling dislocation dynamics and crystal plasticity; (15) New 'importance sampling' methods were developed and demonstrated to reduce the computational cost of rare-event inference; (16) The survey and evaluation of existing data and knowledge bases was updated for NE-KAMS; (17) The NEAMS Early User Program was launched; (18) The Nuclear Regulatory Commission (NRC) Office of Regulatory Research was introduced to the NEAMS program; (19) The NEAMS overall software quality assurance plan (SQAP) was revised to version 1.5; and (20) Work continued on NiCE and its plug-ins and other utilities, such as Cubit and VisIt.« less

Differential Die-Away Instrument: Report on Benchmark Measurements and Comparison with Simulation for the Effects of Neutron Poisons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir

2015-03-30

In this report, new experimental data and MCNPX simulation results of the differential die-away (DDA) instrument response to the presence of neutron absorbers are evaluated. In our previous fresh nuclear fuel experiments and simulations, no neutron absorbers or poisons were included in the fuel definition. These new results showcase the capability of the DDA instrument to acquire data from a system that better mimics spent nuclear fuel.

Comparison of Homogeneous and Heterogeneous CFD Fuel Models for Phase I of the IAEA CRP on HTR Uncertainties Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom; Su-Jong Yoon

2014-04-01

Computational Fluid Dynamics (CFD) evaluation of homogeneous and heterogeneous fuel models was performed as part of the Phase I calculations of the International Atomic Energy Agency (IAEA) Coordinate Research Program (CRP) on High Temperature Reactor (HTR) Uncertainties in Modeling (UAM). This study was focused on the nominal localized stand-alone fuel thermal response, as defined in Ex. I-3 and I-4 of the HTR UAM. The aim of the stand-alone thermal unit-cell simulation is to isolate the effect of material and boundary input uncertainties on a very simplified problem, before propagation of these uncertainties are performed in subsequent coupled neutronics/thermal fluids phasesmore » on the benchmark. In many of the previous studies for high temperature gas cooled reactors, the volume-averaged homogeneous mixture model of a single fuel compact has been applied. In the homogeneous model, the Tristructural Isotropic (TRISO) fuel particles in the fuel compact were not modeled directly and an effective thermal conductivity was employed for the thermo-physical properties of the fuel compact. On the contrary, in the heterogeneous model, the uranium carbide (UCO), inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers of the TRISO fuel particles are explicitly modeled. The fuel compact is modeled as a heterogeneous mixture of TRISO fuel kernels embedded in H-451 matrix graphite. In this study, a steady-state and transient CFD simulations were performed with both homogeneous and heterogeneous models to compare the thermal characteristics. The nominal values of the input parameters are used for this CFD analysis. In a future study, the effects of input uncertainties in the material properties and boundary parameters will be investigated and reported.« less

Investigation of the transient fuel preburner manifold and combustor

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Chen, Yen-Sen; Farmer, Richard C.

1989-01-01

A computational fluid dynamics (CFD) model with finite rate reactions, FDNS, was developed to study the start transient of the Space Shuttle Main Engine (SSME) fuel preburner (FPB). FDNS is a time accurate, pressure based CFD code. An upwind scheme was employed for spatial discretization. The upwind scheme was based on second and fourth order central differencing with adaptive artificial dissipation. A state of the art two-equation k-epsilon (T) turbulence model was employed for the turbulence calculation. A Pade' Rational Solution (PARASOL) chemistry algorithm was coupled with the point implicit procedure. FDNS was benchmarked with three well documented experiments: a confined swirling coaxial jet, a non-reactive ramjet dump combustor, and a reactive ramjet dump combustor. Excellent comparisons were obtained for the benchmark cases. The code was then used to study the start transient of an axisymmetric SSME fuel preburner. Predicted transient operation of the preburner agrees well with experiment. Furthermore, it was also found that an appreciable amount of unburned oxygen entered the turbine stages.

Integral edge seals for phosphoric acid fuel cells

DOEpatents

Granata, Jr., Samuel J.; Woodle, Boyd M.; Dunyak, Thomas J.

1992-01-01

A phosphoric acid fuel cell having integral edge seals formed by an elastomer permeating an outer peripheral band contiguous with the outer peripheral edges of the cathode and anode assemblies and the matrix to form an integral edge seal which is reliable, easy to manufacture and has creep characteristics similar to the anode, cathode and matrix assemblies inboard of the seals to assure good electrical contact throughout the life of the fuel cell.

Vibrating fuel grapple

DOEpatents

Chertock, deceased, Alan J.; Fox, Jack N.; Weissinger, Robert B.

1982-01-01

A reactor refueling method utilizing a vibrating fuel grapple for removing spent fuel assemblies from a reactor core which incorporates a pneumatic vibrator in the grapple head, enabling additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

NASA Astrophysics Data System (ADS)

Schneider, E. A.; Deinert, M. R.; Cady, K. B.

2006-10-01

The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

Measuring the Multiplication of Spent Fuel Assemblies – It’s easier than you think!

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen Joseph

This is a set of eight slides which advertise how easy it can be to measure the multiplication of a spent fuel assembly. A robust (fission chambers), rapid (under 15 minutes), direct (multiplication is measured, not photons from fission fragments) measurement of multiplication is possible.

Validation of the U.S. NRC NGNP evaluation model with the HTTR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saller, T.; Seker, V.; Downar, T.

2012-07-01

The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water reactor (LWR) homogenization methods rely on the short mean free paths of neutrons in LWR. In gas-cooled, graphite-moderated reactors like the HTTR neutrons have much longer mean free paths and penetrate further into neighboring assemblies than in LWRs. Because of this, conventional lattice calculations with a single assembly may not be valid. In addition to difficulties caused by the longer mean free paths, the HTTR presents unique axial and radial heterogeneities that require additional modifications to the single assembly homogenization method. To handle these challenges, the homogenizationmore » domain is decreased while the computational domain is increased. Instead of homogenizing a single hexagonal fuel assembly, the assembly is split into six triangles on the radial plane and five blocks axially in order to account for the placement of burnable poisons. Furthermore, the radial domain is increased beyond a single fuel assembly to account for spectrum effects from neighboring fuel, reflector, and control rod assemblies. A series of five two-dimensional cases, each closer to the full core, were calculated to evaluate the effectiveness of the homogenization method and cross-sections. (authors)« less

Development and applications of retro-reflective surfaces for ultrasound in LBE

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

The Belgian Nuclear Research Centre SCK-CEN is in the process of developing MYRRHA - a fast flux research reactor to replace the aging BR2. MYRRHA is conceptualized as an accelerator driven system cooled by lead bismuth eutectic mixture (LBE). As LBE is opaque to visual light, ultrasonic measurement techniques are employed as the main technology to provide feedback to submerged operations when needed. Conceptually, MYRRHA is a pool type reactor divided in a cold lower plenum and hot upper plenum separated by a diaphragm that forces the main flow through the core. The main flow is cooled by four heatmore » exchangers and driven by two liquid metal pumps. One of the tasks tackled using ultrasound is locating a potentially lost fuel assembly to assist a recovery operation. As all fuel manipulations in MYRRHA are performed in the lower plenum, a potentially lost fuel assembly is located in the lower plenum. Buoyancy will force the lost fuel assembly to float against the diaphragm unless it is still partially inserted in the core. Because of the latter situation, an ultrasonic scan localizing the fuel assembly should be performed from a large distance to avoid a collision with such a partially inserted fuel assembly. Unfortunately, standard machined stainless steel objects, such as a fuel assembly, reflect an ultrasonic pulse in a specular way which induces stringent requirements on the alignment of the ultrasonic sensor with respect to the fuel assembly as we cannot rely on diffuse reflections and/or scattering of the ultrasonic pulse. Moreover, increasing the distance also increases geometric spreading and absorption of the pulse weakening the signal amplitude even faster to noise levels when deviating from perfect alignment. An alternative approach consists in relying on reflections from the known surroundings: a lost fuel assembly will block the line-of-sight to the diaphragm resulting in an anomaly in the reflection - either a shorter than expected time-of-flight of the pulse or a complete absence of a reflection like a shadow. In that way, it suffices to align the sensor with the diaphragm instead of the fuel assembly which is much easier to achieve as the robotics on which the sensor is mounted move parallel with the diaphragm. The alignment requirement in the latter approach can be further relaxed by using a tiling of retro-reflectors on the lower surface of the diaphragm. In that way, alignment becomes less vital and the main source of acoustic energy loss - geometric spread of the beam - is almost completely removed, leaving only absorption losses. In this paper, we present the first results in developing a retro reflectance surface for ultrasound in LBE. We present experimental results for different designs of retro-reflectors in both water and LBE. We discuss both linear and array retro-reflectors of different sizes and investigate the influence of the main relevant ultrasonic parameters such as wavelength and spot size on the strength of the received reflection under different alignment angles. We also demonstrate how retro-reflective surfaces can be exploited when localizing objects using linear and rotating scanning methods. (authors)« less

rnaQUAST: a quality assessment tool for de novo transcriptome assemblies.

PubMed

Bushmanova, Elena; Antipov, Dmitry; Lapidus, Alla; Suvorov, Vladimir; Prjibelski, Andrey D

2016-07-15

Ability to generate large RNA-Seq datasets created a demand for both de novo and reference-based transcriptome assemblers. However, while many transcriptome assemblers are now available, there is still no unified quality assessment tool for RNA-Seq assemblies. We present rnaQUAST-a tool for evaluating RNA-Seq assembly quality and benchmarking transcriptome assemblers using reference genome and gene database. rnaQUAST calculates various metrics that demonstrate completeness and correctness levels of the assembled transcripts, and outputs them in a user-friendly report. rnaQUAST is implemented in Python and is freely available at http://bioinf.spbau.ru/en/rnaquast ap@bioinf.spbau.ru Supplementary data are available at Bioinformatics online. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com.

Little Boy replication: justification and construction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Malenfant, R.E.

A reconstruction of the Little Boy weapon allowed experiments to evaluate yield, leakage measurements for comparison with calculations, and phenomenological measurements to evaluate various in-situ dosimeters. The reconstructed weapon was operated at sustained delayed critical at the Los Alamos Critical Assembly Facility. The present experiments provide a wealth of information to benchmark calculations and demonstrate that the 1965 measurements on the Ichiban assembly (a spherical mockup of Little Boy) were in error.

Electrochemical Characterization Laboratory | Energy Systems Integration

Science.gov Websites

proton exchange membrane fuel cells. Photo of an NREL researcher evaluating catalyst activity in the the following capabilities: Determination and benchmarking of novel electrocatalyst activity

Single molecule sequencing-guided scaffolding and correction of draft assemblies.

PubMed

Zhu, Shenglong; Chen, Danny Z; Emrich, Scott J

2017-12-06

Although single molecule sequencing is still improving, the lengths of the generated sequences are inevitably an advantage in genome assembly. Prior work that utilizes long reads to conduct genome assembly has mostly focused on correcting sequencing errors and improving contiguity of de novo assemblies. We propose a disassembling-reassembling approach for both correcting structural errors in the draft assembly and scaffolding a target assembly based on error-corrected single molecule sequences. To achieve this goal, we formulate a maximum alternating path cover problem. We prove that this problem is NP-hard, and solve it by a 2-approximation algorithm. Our experimental results show that our approach can improve the structural correctness of target assemblies in the cost of some contiguity, even with smaller amounts of long reads. In addition, our reassembling process can also serve as a competitive scaffolder relative to well-established assembly benchmarks.

Fibre optic sensor based measurements of flow-induced vibration in a liquid metal cooled nuclear reactor set-up

NASA Astrophysics Data System (ADS)

De Pauw, B.; Vanlanduit, S.; Van Tichelen, K.; Geernaert, T.; Thienpont, H.; Berghmans, F.

2017-04-01

Fuel assembly vibrations in nuclear reactor cores should not be excessive as these can compromise the lifetime of the assembly and lead to safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants. We therefore demonstrate accurate measurements of the vibrations of a fuel assembly in a lead-bismuth eutectic cooled installation with fibre Bragg grating (FBG) based sensors. The use of FBGs in combination with a dedicated sensor integration approach allows accounting for the severe geometrical constraints and providing for the required minimal intrusiveness of the instrumentation, identifying the vibration modes with required accuracy and observing the differences between the vibration amplitudes of the individual fuel pins as well as evidencing a low frequency fuel pin vibration mode resulting from the supports.

Consistent criticality and radiation studies of Swiss spent nuclear fuel: The CS2M approach.

PubMed

Rochman, D; Vasiliev, A; Ferroukhi, H; Pecchia, M

2018-06-15

In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS 2 M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO 2 assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources. Copyright © 2018 Elsevier B.V. All rights reserved.

Closeup view of the bottom area of Space Shuttle Main ...

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

Close-up view of the bottom area of Space Shuttle Main Engine (SSME) 2052 engine assembly mounted in a SSME Engine Handler in the Horizontal Processing area of the SSME Processing Facility at Kennedy Space Center. The most prominent features in this view are the Low-Pressure Oxidizer Discharge Duct toward the bottom of the assembly, the SSME Engine Controller and the Main Fuel Valve Hydraulic Actuator are in the approximate center of the assembly in this view, the Low-Pressure Fuel Turbopump (LPFTP), the LPFTP Discharge Duct are to the left on the assembly in this view and the High-Pressure Fuel Turbopump is located toward the top of the engine assembly in this view. The ring of tabs in the right side of the image, at the approximate location of the Nozzle and the Coolant Outlet Manifold interface is the Heat Shield Support Ring. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

NASA Technical Reports Server (NTRS)

Rohal, R. G.; Tambling, T. N.; Smith, R. L.

1973-01-01

Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

Gamma and fast neutron radiation monitoring inside spent reactor fuel assemblies

NASA Astrophysics Data System (ADS)

Lakosi, L.; Tam Nguyen, C.

2007-09-01

Gamma and neutron signatures of spent reactor fuel were monitored by small-size silicon diode and track etch detectors, respectively, in a nuclear power plant (NPP). These signatures, reflecting gross gamma intensity and the 242,244Cm content, contain information on the burn-up (BU) and cooling time (CT) of the fuel. The small size of the detectors allows close access to inside parts of the assemblies out of reach of other methods. A commercial Si diode was encapsulated in a cylindrical steel case and was used for gross γ monitoring. CR-39 detectors were used for neutron measurements. Irradiation exposures at the NPP were implemented in the central dosimetric channel of spent fuel assemblies (SFAs) stored in borated water. Gross γ and neutron axial profiles were taken up by scanning with the aid of a long steel guide tube, lowered down to the spent fuel pond by crane and fitted to the headpiece of the fuel assemblies. Gamma measurements were performed using a long cable introduced in this tube, with the Si diode at the end. A long steel wire was also led through the guide tube, to which a chain of 15 sample holder capsules was attached, each containing a track detector. Gamma dose rates of 0.1-10 kGy h -1, while neutron fluxes in a range of (0.25-26) 10 4 cm -2 s -1 were recorded. The results are in good correlation with those of a calculation for spent fuel neutron yield.

Outcomes of the JNT 1955 Phase I Viability Study of Gamma Emission Tomography for Spent Fuel Verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacobsson-Svard, Staffan; Smith, Leon E.; White, Timothy

The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly has been assessed within the IAEA Support Program project JNT 1955, phase I, which was completed and reported to the IAEA in October 2016. Two safeguards verification objectives were identified in the project; (1) independent determination of the number of active pins that are present in a measured assembly, in the absence of a priori information about the assembly; and (2) quantitative assessment of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup,more » under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives was evaluated across a range of fuel types, burnups and cooling times, while targeting a total interrogation time of less than 60 minutes. The evaluations were founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. The simulated instrument response data were then processed using a variety of tomographic-reconstruction and image-processing methods, and scoring metrics were defined and used to evaluate the performance of the methods.This paper describes the analysis framework and metrics used to predict tomographer performance. It also presents the design of a “universal” GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA. Finally, it gives examples of the expected partial-defect detection capabilities for some fuels and diversion scenarios, and it provides a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.« less

Saturn Apollo Program

NASA Image and Video Library

1964-09-01

This photograph shows the components for the Saturn V S-IC stage fuel tank assembly in the Manufacturing Engineering Laboratory, building 4707, at the Marshall Space Flight Center (MSFC). Left to right are upper head, lower head, and forward skirt assembly. Thirty-three feet in diameter, they will hold a total of 4,400,000 pounds of fuel. Although this tankage was assembled at MSFC, the elements were made by the Boeing Company at Wichita and the Michould Operations at New Orleans.

Cooling assembly for fuel cells

DOEpatents

Kaufman, Arthur; Werth, John

1990-01-01

A cooling assembly for fuel cells having a simplified construction whereby coolant is efficiently circulated through a conduit arranged in serpentine fashion in a channel within a member of such assembly. The channel is adapted to cradle a flexible, chemically inert, conformable conduit capable of manipulation into a variety of cooling patterns without crimping or otherwise restricting of coolant flow. The conduit, when assembled with the member, conforms into intimate contact with the member for good thermal conductivity. The conduit is non-corrodible and can be constructed as a single, manifold-free, continuous coolant passage means having only one inlet and one outlet.

System Regulates the Water Contents of Fuel-Cell Streams

NASA Technical Reports Server (NTRS)

Vasquez, Arturo; Lazaroff, Scott

2005-01-01

An assembly of devices provides for both humidification of the reactant gas streams of a fuel cell and removal of the product water (the water generated by operation of the fuel cell). The assembly includes externally-sensing forward-pressure regulators that supply reactant gases (fuel and oxygen) at variable pressures to ejector reactant pumps. The ejector supply pressures depend on the consumption flows. The ejectors develop differential pressures approximately proportional to the consumption flow rates at constant system pressure and with constant flow restriction between the mixer-outlet and suction ports of the ejectors. For removal of product water from the circulating oxygen stream, the assembly includes a water/gas separator that contains hydrophobic and hydrophilic membranes. The water separator imposes an approximately constant flow restriction, regardless of the quality of the two-phase flow that enters it from the fuel cell. The gas leaving the water separator is nearly 100 percent humid. This gas is returned to the inlet of the fuel cell along with a quantity of dry incoming oxygen, via the oxygen ejector, thereby providing some humidification.

RERTR-12 Post-irradiation Examination Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, Francine; Williams, Walter; Robinson, Adam

2015-02-01

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-propertymore » changes of the fuel and cladding.« less

Apparatus for inspecting fuel elements

DOEpatents

Oakley, David J.; Groves, Oliver J.; Kaiser, Bruce J.

1986-01-01

Disclosed is an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

Apparatus for inspecting fuel elements

DOEpatents

Kaiser, B.J.; Oakley, D.J.; Groves, O.J.

1984-12-21

This disclosure describes an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

Electrolyte creepage barrier for liquid electrolyte fuel cells

DOEpatents

Li, Jian [Alberta, CA; Farooque, Mohammad [Danbury, CT; Yuh, Chao-Yi [New Milford, CT

2008-01-22

A dielectric assembly for electrically insulating a manifold or other component from a liquid electrolyte fuel cell stack wherein the dielectric assembly includes a substantially impermeable dielectric member over which electrolyte is able to flow and a barrier adjacent the dielectric member and having a porosity of less than 50% and greater than 10% so that the barrier is able to measurably absorb and chemically react with the liquid electrolyte flowing on the dielectric member to form solid products which are stable in the liquid electrolyte. In this way, the barrier inhibits flow or creepage of electrolyte from the dielectric member to the manifold or component to be electrically insulated from the fuel cell stack by the dielectric assembly.

CFD Analysis of Coolant Flow in VVER-440 Fuel Assemblies with the Code ANSYS CFX 10.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toth, Sandor; Legradi, Gabor; Aszodi, Attila

2006-07-01

From the aspect of planning the power upgrading of nuclear reactors - including the VVER-440 type reactor - it is essential to get to know the flow field in the fuel assembly. For this purpose we have developed models of the fuel assembly of the VVER-440 reactor using the ANSYS CFX 10.0 CFD code. At first a 240 mm long part of a 60 degrees segment of the fuel pin bundle was modelled. Implementing this model a sensitivity study on the appropriate meshing was performed. Based on the development of the above described model, further models were developed: a 960more » mm long part of a 60-degree-segment and a full length part (2420 mm) of the fuel pin bundle segment. The calculations were run using constant coolant properties and several turbulence models. The impacts of choosing different turbulence models were investigated. The results of the above-mentioned investigations are presented in this paper. (authors)« less

DIANA: A multi-phase, multi-component hydrodynamic model for the analysis of severe accidents in heavy water reactors with multiple-tube assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tentner, A.M.

1994-03-01

A detailed hydrodynamic fuel relocation model has been developed for the analysis of severe accidents in Heavy Water Reactors with multiple-tube Assemblies. This model describes the Fuel Disruption and Relocation inside a nuclear fuel assembly and is designated by the acronym DIANA. DIANA solves the transient hydrodynamic equations for all the moving materials in the core and treats all the relevant flow regimes. The numerical solution techniques and some of the physical models included in DIANA have been developed taking advantage of the extensive experience accumulated in the development and validation of the LEVITATE (1) fuel relocation model of SAS4Amore » [2, 3]. The model is designed to handle the fuel and cladding relocation in both voided and partially voided channels. It is able to treat a wide range of thermal/ hydraulic/neutronic conditions and the presence of various flow regimes at different axial locations within the same hydrodynamic channel.« less

Fuel cell assembly with electrolyte transport

DOEpatents

Chi, Chang V.

1983-01-01

A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

MELCOR model for an experimental 17x17 spent fuel PWR assembly.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardoni, Jeffrey

2010-11-01

A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru; Pinegin, A. A.

The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

Vibrating fuel grapple. [LMFBR

DOEpatents

Chertock, A.J.; Fox, J.N.; Weissinger, R.B.

A reactor refueling method is described which utilizes a vibrating fuel grapple for removing spent fuel assemblies from a reactor core. It incorporates a pneumatic vibrator in the grapple head which allows additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

TMI-2 (Three Mile Island Unit 2) core region defueling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rodabaugh, J.M.; Cowser, D.K.

1988-01-01

In July of 1982, a video camera was inserted into the Three Mile Island Unit 2 reactor vessel providing the first visual evidence of core damage. This inspection, and numerous subsequent data acquisition tasks, revealed a central void /approx/1.5 m (5 ft) deep. This void region was surrounded by partial length fuel assemblies and ringed on the periphery by /approx/40 full-length, but partial cross-section, fuel assemblies. All of the original 177 fuel assemblies exhibited signs of damage. The bottom of the void cavity was covered with a bed of granular rubble, fuel assembly upper end fittings, control rod spiders, fuelmore » rod fragments, and fuel pellets. It was obvious that the normal plant refueling system not suitable for removing the damaged core. A new system of defueling tools and equipment was necessary to perform this task. Design of the new system was started immediately, followed by >1 yr of fabrication. Delivery and checkout of the defueling system occurred in mid-1985. Actual defueling was initiated in late 1985 with removal of the debris bed at the bottom of the core void. Obstructions to the debris, such as end fittings and fuel rod fragments ere removed first; then /approx/23,000 kg (50,000lb) of granular debris was quickly loaded into canisters. Core region defueling was completed in late 1987, /approx/2 yr after it was initiated.« less

Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

DOE Office of Scientific and Technical Information (OSTI.GOV)

Patra, Anirban; Tomé, Carlos N.

A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

DOE PAGES

Patra, Anirban; Tomé, Carlos N.

2017-03-06

A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

Importance of balancing membrane and electrode water in anion exchange membrane fuel cells

NASA Astrophysics Data System (ADS)

Omasta, T. J.; Wang, L.; Peng, X.; Lewis, C. A.; Varcoe, J. R.; Mustain, W. E.

2018-01-01

Anion exchange membrane fuel cells (AEMFCs) offer several potential advantages over proton exchange membrane fuel cells (PEMFCs), most notably to overcome the cost barrier that has slowed the growth and large scale implementation of fuel cells for transportation. However, limitations in performance have held back AEMFCs, specifically in the areas of stability, carbonation, and maximum achievable current and power densities. In order for AEMFCs to contend with PEMFCs for market viability, it is necessary to realize a competitive cell performance. This work demonstrates a new benchmark for a H2/O2 AEMFC with a peak power density of 1.4 W cm-2 at 60 °C. This was accomplished by taking a more precise look at balancing necessary membrane hydration while preventing electrode flooding, which somewhat surprisingly can occur both at the anode and the cathode. Specifically, radiation-grafted ETFE-based anion exchange membranes and anion exchange ionomer powder, functionalized with benchmark benzyltrimethylammonium groups, were utilized to examine the effects of the following parameters on AEMFC performance: feed gas flow rate, the use of hydrophobic vs. hydrophilic gas diffusion layers, and gas feed dew points.

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, M. A.; Sen, R. S.; Ougouag, A. M.

2012-07-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space availablemore » for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)« less

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag

2012-04-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel,more » the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Storey, Robson, F.; Mauritz, Kenneth, A.; Patton, Derek, L.

The overall objective of this project was the development and evaluation of novel hydrocarbon fuel cell (FC) membranes that possess high temperature performance and long term chemical/mechanical durability in proton exchange membrane (PEM) fuel cells (FC). The major research theme was synthesis of aromatic hydrocarbon polymers of the poly(arylene ether sulfone) (PAES) type containing sulfonic acid groups tethered to the backbone via perfluorinated alkylene linkages and in some cases also directly attached to the phenylene groups along the backbone. Other research themes were the use of nitrogen-based heterocyclics instead of acid groups for proton conduction, which provides high temperature, lowmore » relative humidity membranes with high mechanical/thermal/chemical stability and pendant moieties that exhibit high proton conductivities in the absence of water, and synthesis of block copolymers consisting of a proton conducting block coupled to poly(perfluorinated propylene oxide) (PFPO) blocks. Accomplishments of the project were as follows: 1) establishment of a vertically integrated program of synthesis, characterization, and evaluation of FC membranes, 2) establishment of benchmark membrane performance data based on Nafion for comparison to experimental membrane performance, 3) development of a new perfluoroalkyl sulfonate monomer, N,N-diisopropylethylammonium 2,2-bis(p-hydroxyphenyl) pentafluoropropanesulfonate (HPPS), 4) synthesis of random and block copolymer membranes from HPPS, 5) synthesis of block copolymer membranes containing high-acid-concentration hydrophilic blocks consisting of HPPS and 3,3'-disulfonate-4,4'-dichlorodiphenylsulfone (sDCDPS), 6) development of synthetic routes to aromatic polymer backbones containing pendent 1H-1,2,3-triazole moieties, 7) development of coupling strategies to create phase-separated block copolymers between hydrophilic sulfonated prepolymers and commodity polymers such as PFPO, 8) establishment of basic performance properties of experimental membranes, 9) fabrication and FC performance testing of membrane electrode assemblies (MEA) from experimental membranes, and 10) measurement of ex situ and in situ membrane durability of experimental membranes. Although none of the experimental hydrocarbon membranes that issued from the project displayed proton conductivities that met DOE requirements, the project contributed to our basic understanding of membrane structure-property relationships in a number of key respects. An important finding of the benchmark studies is that physical degradation associated with humidity and temperature variations in the FC tend to open new fuel crossover pathways and act synergistically with chemical degradation to accelerate overall membrane degradation. Thus, for long term membrane survival and efficient fuel utilization, membranes must withstand internal stresses due to humidity and temperature changes. In this respect, rigid aromatic hydrocarbon fuel cell membranes, e.g. PAES, offer an advantage over un-modified Nafion membranes. The benchmark studies also showed that broadband dielectric spectroscopy is a potentially powerful tool in assessing shifts in the fundamental macromolecular dynamics caused by Nafion chemical degradation, and thus, this technique is of relevance in interrogating proton exchange membrane durability in fuel cells and macromolecular dynamics as coupled to proton migration, which is of fundamental relevance in proton exchange membranes in fuel cells. A key finding from the hydrocarbon membrane synthesis effort was that rigid aromatic polymers containing isolated ion exchange groups tethered tightly to the backbone (short tether), such as HPPS, provide excellent mechanical and durability properties but do not provide sufficient conductivity, in either random or block configuration, when used as the sole ion exchange monomer. However, we continue to hypothesize that longer tethers, and tethered groups spaced more closely within the hydrophilic chain elements of the polymer, will yield highly conductive materials with excellent mechanical properties. Another key finding is the superior performance of PAES membranes upon being subjected to open circuit voltage (OCV) testing. Throughout the course of the experiment, OCV for the PAES not only stayed higher but also decayed at a much lower rate, which is attributed to better dimensional stability and improved mechanical and gas barrier properties. The rigid backbone reinforcement of PAES adds gas diffusion tortuosity that restricts membrane degradation and OCV loss due to reduced fuel crossover. The overall results of creep, contractile stress and mechanical tensile tests confirm the conclusion that degraded MEAs of PAES membrane can handle stress and are more likely to be more durable in a fuel cell, even after subjected to 62h of OCV degradation.« less

QUAST: quality assessment tool for genome assemblies.

PubMed

Gurevich, Alexey; Saveliev, Vladislav; Vyahhi, Nikolay; Tesler, Glenn

2013-04-15

Limitations of genome sequencing techniques have led to dozens of assembly algorithms, none of which is perfect. A number of methods for comparing assemblers have been developed, but none is yet a recognized benchmark. Further, most existing methods for comparing assemblies are only applicable to new assemblies of finished genomes; the problem of evaluating assemblies of previously unsequenced species has not been adequately considered. Here, we present QUAST-a quality assessment tool for evaluating and comparing genome assemblies. This tool improves on leading assembly comparison software with new ideas and quality metrics. QUAST can evaluate assemblies both with a reference genome, as well as without a reference. QUAST produces many reports, summary tables and plots to help scientists in their research and in their publications. In this study, we used QUAST to compare several genome assemblers on three datasets. QUAST tables and plots for all of them are available in the Supplementary Material, and interactive versions of these reports are on the QUAST website. http://bioinf.spbau.ru/quast . Supplementary data are available at Bioinformatics online.

Chalcogen catalysts for polymer electrolyte fuel cell

DOEpatents

Alonso-Vante, Nicolas [Buxerolles, FR; Zelenay, Piotr [Los Alamos, NM; Choi, Jong-Ho [Los Alamos, NM; Wieckowski, Andrzej [Champaign, IL; Cao, Dianxue [Urbana, IL

2009-09-15

A methanol-tolerant cathode catalyst and a membrane electrode assembly for fuel cells that includes such a cathode catalyst. The cathode catalyst includes a support having at least one transition metal in elemental form and a chalcogen disposed on the support. Methods of making the cathode catalyst and membrane electrode assembly are also described.

Mixer Assembly for a Gas Turbine Engine

NASA Technical Reports Server (NTRS)

Smith, Lance L. (Inventor); Fotache, Catalin G. (Inventor); Dai, Zhongtao (Inventor); Cohen, Jeffrey M. (Inventor); Hautman, Donald J. (Inventor)

2015-01-01

A mixer assembly for a gas turbine engine is provided, including a main mixer with fuel injection holes located between at least one radial swirler and at least one axial swirler, wherein the fuel injected into the main mixer is atomized and dispersed by the air flowing through the radial swirler and the axial swirler.

Mixer Assembly for a Gas Turbine Engine

NASA Technical Reports Server (NTRS)

Dai, Zhongtao (Inventor); Cohen, Jeffrey M. (Inventor); Fotache, Catalin G. (Inventor); Hautman, Donald J. (Inventor); Smith, Lance L. (Inventor)

2018-01-01

A mixer assembly for a gas turbine engine is provided, including a main mixer with fuel injection holes located between at least one radial swirler and at least one axial swirler, wherein the fuel injected into the main mixer is atomized and dispersed by the air flowing through the radial swirler and the axial swirler.

Chalcogen catalysts for polymer electrolyte fuel cell

DOEpatents

Zelenay, Piotr; Choi, Jong-Ho; Alonso-Vante, Nicolas; Wieckowski, Andrzej; Cao, Dianxue

2010-08-24

A methanol-tolerant cathode catalyst and a membrane electrode assembly for fuel cells that includes such a cathode catalyst. The cathode catalyst includes a support having at least one transition metal in elemental form and a chalcogen disposed on the support. Methods of making the cathode catalyst and membrane electrode assembly are also described.

Method of making MEA for PEM/SPE fuel cell

DOEpatents

Hulett, Jay S.

2000-01-01

A method of making a membrane-electrode-assembly (MEA) for a PEM/SPE fuel cell comprising applying a slurry of electrode-forming material directly onto a membrane-electrolyte film. The slurry comprises a liquid vehicle carrying catalyst particles and a binder for the catalyst particles. The membrane-electrolyte is preswollen by contact with the vehicle before the electrode-forming slurry is applied to the membrane-electrolyte. The swollen membrane-electrolyte is constrained against shrinking in the "x" and "y" directions during drying. Following assembly of the fuel cell, the MEA is rehydrated inside the fuel cell such that it swells in the "z" direction for enhanced electrical contact with contiguous electrically conductive components of the fuel cell.

Detachable connection for a nuclear reactor fuel assembly

DOEpatents

Christiansen, D.W.; Karnesky, R.A.

1983-08-29

A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

Process for recycling components of a PEM fuel cell membrane electrode assembly

DOEpatents

Shore, Lawrence [Edison, NJ

2012-02-28

The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

Detachable connection for a nuclear reactor fuel assembly

DOEpatents

Christiansen, David W.; Karnesky, Richard A.

1986-01-01

A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engageable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, Richard L.; Folsom, Charles Pearson; Pastore, Giovanni

2016-05-01

One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

Little Boy replication: justification and construction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Malenfant, R.E.

A reconstruction of the Little Boy weapon allowed experiments to evaluate yield, leakage measurements for comparison with calculations, and phenomenological measurements to evaluate various in-situ dosimeters. The reconstructed weapon was operated at sustained delayed critical at the Los Alamos Critical Assembly Facility. The present experiments provide a wealth of information to benchmark calculations and demonstrate that the 1965 measurements on the Ichiban assembly (a spherical mockup of Little Boy) were in error. 5 references, 2 figures.

Measuring spent fuel assembly multiplication in borated water with a passive neutron albedo reactivity instrument

NASA Astrophysics Data System (ADS)

Tobin, Stephen J.; Peura, Pauli; Bélanger-Champagne, Camille; Moring, Mikael; Dendooven, Peter; Honkamaa, Tapani

2018-07-01

The performance of a passive neutron albedo reactivity (PNAR) instrument to measure neutron multiplication of spent nuclear fuel in borated water is investigated as part of an integrated non-destructive assay safeguards system. To measure the PNAR Ratio, which is proportional to the neutron multiplication, the total neutron count rate is measured in high- and low-multiplying environments by the PNAR instrument. The integrated system also contains a load cell and a passive gamma emission tomograph, and as such meets all the recommendations of the IAEA's recent ASTOR Experts Group report. A virtual spent fuel library for VVER-440 fuel was used in conjunction with MCNP simulations of the PNAR instrument to estimate the measurement uncertainties from (1) variation in the water boron content, (2) assembly positioning in the detector and (3) counting statistics. The estimated aggregate measurement uncertainty on the PNAR Ratio measurement is 0.008, to put this uncertainty in context, the difference in the PNAR Ratio between a fully irradiated assembly and this same assembly when fissile isotopes only absorb neutrons, but do not emit neutrons, is 0.106, a 13-sigma effect. The 1-sigma variation of 0.008 in the PNAR Ratio is estimated to correspond to a 3.2 GWd/tU change in assembly burnup.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, Charles W.

1998-01-01

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, C.W.

1998-11-03

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

NNSA B-Roll: MOX Facility

ScienceCinema

None

2017-12-09

In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

NNSA B-Roll: MOX Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2010-05-21

In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

All About MOX

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2009-07-29

In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

FUEL ASSEMBLY FOR A NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1958-04-29

A fuel assembly for a nuclear reactor of the type wherein liquid coolant is circulated through the core of the reactor in contact with the external surface of the fuel elements is described. In this design a plurality of parallel plates containing fissionable material are spaced about one-tenth of an inch apart and are supported between a pair of spaced parallel side members generally perpendicular to the plates. The plates all have a small continuous and equal curvature in the same direction between the side members.

All About MOX

ScienceCinema

None

2018-01-16

In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

Fabrication Method for Laboratory-Scale High-Performance Membrane Electrode Assemblies for Fuel Cells.

PubMed

Sassin, Megan B; Garsany, Yannick; Gould, Benjamin D; Swider-Lyons, Karen E

2017-01-03

Custom catalyst-coated membranes (CCMs) and membrane electrode assemblies (MEAs) are necessary for the evaluation of advanced electrocatalysts, gas diffusion media (GDM), ionomers, polymer electrolyte membranes (PEMs), and electrode structures designed for use in next-generation fuel cells, electrolyzers, or flow batteries. This Feature provides a reliable and reproducible fabrication protocol for laboratory scale (10 cm 2 ) fuel cells based on ultrasonic spray deposition of a standard Pt/carbon electrocatalyst directly onto a perfluorosulfonic acid PEM.

Fuel inspection and reconstitution experience at Surry Power Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brookmire, T.A.

Surry Power Station, located on the James River near Williamsburg, Virginia, has two Westinghouse pressurized water reactors. Unit 2 consistently sets a high standard of fuel performance (no indication of fuel failures in recent cycles), while unit 1, since cycle 6, has been plagued with numerous fuel failures. Both Surry units operate with Westinghouse standard 15 x 15 fuel. Virginia Power management set goals to reduce the coolant activity, thus reducing person-rem exposure and the associated costs of high coolant activity. To achieve this goal, extensive fuel examination campaigns were undertaken that included high-magnification video inspectionsa, debris cleaning, wet andmore » vacuum fuel sipping, fuel rod ultrasonic testing, and eddy current examination. In the summer of 1985, during cycle 8 operation, Kraftwerk Union reconstituted (repaired) the damage, once-burned assemblies from cycles 6 and 7 by replacing failed fuel rods with solid Zircaloy-4 rods. Currently, cycle 9 has operated for 5 months without any indication of fuel failure (the cycle 9 core has two reconstituted assemblies).« less

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

DOE PAGES

Carmack, W. Jon; Chichester, Heather M.; Porter, Douglas L.; ...

2016-02-27

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This then places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. After comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

Analysis of key safety metrics of thorium utilization in LWRs

DOE PAGES

Ade, Brian J.; Bowman, Stephen M.; Worrall, Andrew; ...

2016-04-08

Here, thorium has great potential to stretch nuclear fuel reserves because of its natural abundance and because it is possible to breed the 232Th isotope into a fissile fuel ( 233U). Various scenarios exist for utilization of thorium in the nuclear fuel cycle, including use in different nuclear reactor types (e.g., light water, high-temperature gas-cooled, fast spectrum sodium, and molten salt reactors), along with use in advanced accelerator-driven systems and even in fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based onmore » concepts that mix thorium with uranium (UO 2 + ThO 2) or that add fertile thorium (ThO 2) fuel pins to typical LWR fuel assemblies. Utilization of mixed fuel assemblies (PuO 2 + ThO 2) is also possible. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts to the nuclear fuel. Thorium and its irradiation products have different nuclear characteristics from those of uranium and its irradiation products. ThO 2, alone or mixed with UO 2 fuel, leads to different chemical and physical properties of the fuel. These key reactor safety–related issues have been studied at Oak Ridge National Laboratory and documented in “Safety and Regulatory Issues of the Thorium Fuel Cycle” (NUREG/CR-7176, U.S. Nuclear Regulatory Commission, 2014). Various reactor analyses were performed using the SCALE code system for comparison of key performance parameters of both ThO 2 + UO 2 and ThO 2 + PuO 2 against those of UO 2 and typical UO 2 + PuO 2 mixed oxide fuels, including reactivity coefficients and power sharing between surrounding UO 2 assemblies and the assembly of interest. The decay heat and radiological source terms for spent fuel after its discharge from the reactor are also presented. Based on this evaluation, potential impacts on safety requirements and identification of knowledge gaps that require additional analysis or research to develop a technical basis for the licensing of thorium fuel are identified.« less

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, W. J.; Chichester, H. M.; Porter, D. L.

2016-05-01

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

L-Area STS MTR/NRU/NRX Grapple Assembly Closure Mechanics Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huizenga, D. J.

2016-06-08

A review of the closure mechanics associated with the Shielded Transfer System (STS) MTR/NRU/NRX grapple assembly utilized at the Savannah River Site (SRS) was performed. This review was prompted by an operational event which occurred at the Canadian Nuclear Laboratories (CNL) utilizing a DTS-XL grapple assembly which is essentially identical to the STS MTR/NRU/NRX grapple assembly used at the SRS. The CNL operational event occurred when a NRU/NRX fuel basket containing spent nuclear fuel assemblies was inadvertently released by the DTS-XL grapple assembly during a transfer. The SM review of the STS MTR/NRU/NRX grapple assembly will examine the operational aspectsmore » of the STS and the engineered features of the STS which prevent such an event at the SRS. The design requirements for the STS NRU/NRX modifications and the overall layout of the STS are provided in other documents.« less

Proton-Fueled, Reversible DNA Hybridization Chain Assembly for pH Sensing and Imaging.

PubMed

Liu, Lan; Liu, Jin-Wen; Huang, Zhi-Mei; Wu, Han; Li, Na; Tang, Li-Juan; Jiang, Jian-Hui

2017-07-05

Design of DNA self-assembly with reversible responsiveness to external stimuli is of great interest for diverse applications. We for the first time develop a pH-responsive, fully reversible hybridization chain reaction (HCR) assembly that allows sensitive sensing and imaging of pH in living cells. Our design relies on the triplex forming sequences that form DNA triplex with toehold regions under acidic conditions and then induce a cascade of strand displacement and DNA assembly. The HCR assembly has shown dynamic responses in physiological pH ranges with excellent reversibility and demonstrated the potential for in vitro detection and live-cell imaging of pH. Moreover, this method affords HCR assemblies with highly localized fluorescence responses, offering advantages of improving sensitivity and better selectivity. The proton-fueled, reversible HCR assembly may provide a useful approach for pH-related cell biology study and disease diagnostics.

Integral gas seal for fuel cell gas distribution assemblies and method of fabrication

DOEpatents

Dettling, Charles J.; Terry, Peter L.

1985-03-19

A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

Method of fabricating an integral gas seal for fuel cell gas distribution assemblies

DOEpatents

Dettling, Charles J.; Terry, Peter L.

1988-03-22

A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

NASA Astrophysics Data System (ADS)

Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

2018-01-01

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

DOE PAGES

Bess, John D.; Montierth, Leland; Köberl, Oliver; ...

2014-10-09

Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the ²³⁵U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data aremore » greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

NASA Astrophysics Data System (ADS)

Lubina, A. S.; Subbotin, A. S.; Sedov, A. A.; Frolov, A. A.

2016-12-01

The fast sodium reactor fuel assembly (FA) with U-Pu-Zr metallic fuel is described. In comparison with a "classical" fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dawson, D.M.; Guerra, G.; Neider, T.

1995-12-01

This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity ormore » other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose.« less

PBF Reactor Building (PER620). Detail of fuel test assembly in ...

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

PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Towards neat methanol operation of direct methanol fuel cells: a novel self-assembled proton exchange membrane.

PubMed

Li, Jing; Cai, Weiwei; Ma, Liying; Zhang, Yunfeng; Chen, Zhangxian; Cheng, Hansong

2015-04-18

We report here a novel proton exchange membrane with remarkably high methanol-permeation resistivity and excellent proton conductivity enabled by carefully designed self-assembled ionic conductive channels. A direct methanol fuel cell utilizing the membrane performs well with a 20 M methanol solution, very close to the concentration of neat methanol.

A simulation study of thinning and fuel treatments on a wildland-urban interface in eastern Oregon, USA

Treesearch

Alan A. Ager; Andrew J. McMahan; James J. Barrett; Charles W. McHugh

2007-01-01

We simulated long-term forest management activities on 16,000-ha wildland-urban interface in the Blue Mountains near La Grande, Oregon. The study area is targeted for thinning and fuels treatments on both private and Federally managed lands to address forest health and sustainability concerns and reduce the risk of severe wildfire. We modeled number of benchmark...

Development and Application of a Sample Holder for In Situ Gaseous TEM Studies of Membrane Electrode Assemblies for Polymer Electrolyte Fuel Cells.

PubMed

Kamino, Takeo; Yaguchi, Toshie; Shimizu, Takahiro

2017-10-01

Polymer electrolyte fuel cells hold great potential for stationary and mobile applications due to high power density and low operating temperature. However, the structural changes during electrochemical reactions are not well understood. In this article, we detail the development of the sample holder equipped with gas injectors and electric conductors and its application to a membrane electrode assembly of a polymer electrolyte fuel cell. Hydrogen and oxygen gases were simultaneously sprayed on the surfaces of the anode and cathode catalysts of the membrane electrode assembly sample, respectively, and observation of the structural changes in the catalysts were simultaneously carried out along with measurement of the generated voltages.

Integral experiments on thorium assemblies with D-T neutron source

NASA Astrophysics Data System (ADS)

Liu, Rong; Yang, Yiwei; Feng, Song; Zheng, Lei; Lai, Caifeng; Lu, Xinxin; Wang, Mei; Jiang, Li

2017-09-01

To validate nuclear data and code in the neutronics design of a hybrid reactor with thorium, integral experiments in two kinds of benchmark thorium assemblies with a D-T fusion neutron source have been performed. The one kind of 1D assemblies consists of polyethylene and depleted uranium shells. The other kind of 2D assemblies consists of three thorium oxide cylinders. The capture reaction rates, fission reaction rates, and (n, 2n) reaction rates in 232Th in the assemblies are measured by ThO2 foils. The leakage neutron spectra from the ThO2 cylinders are measured by a liquid scintillation detector. The experimental uncertainties in all the results are analyzed. The measured results are compared to the calculated ones with MCNP code and ENDF/B-VII.0 library data.

Comparative TEA for Indirect Liquefaction Pathways to Distillate-Range Fuels via Oxygenated Intermediates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tan, Eric; Snowden-Swan, Lesley J.; Talmadge, Michael

This paper presents a comparative techno-economic analysis of five conversion pathways from biomass to gasoline-, jet-, and diesel-range hydrocarbons via indirect liquefaction with specific focus on pathways utilizing oxygenated intermediates (derived either via thermochemical or biochemical conversion steps). The four emerging pathways of interest are compared with one conventional pathway (Fischer-Tropsch) for the production of the hydrocarbon blendstocks. The processing steps of the four emerging pathways include: biomass-to-syngas via indirect gasification, gas cleanup, conversion of syngas to alcohols/oxygenates, followed by conversion of alcohols/oxygenates to hydrocarbon blendstocks via dehydration, oligomerization, and hydrogenation. We show that the emerging pathways via oxygenated intermediatesmore » have the potential to be cost competitive with the conventional Fischer-Tropsch process. The evaluated pathways and the benchmark process generally exhibit similar fuel yields and carbon conversion efficiencies. The resulting minimum fuel selling prices are comparable to the benchmark at approximately $3.60 per gallon-gasoline equivalent, with potential for two new pathways to be more economically competitive. Additionally, the coproduct values can play an important role in the economics of the processes with oxygenated intermediates derived via syngas fermentation. Major cost drivers for the integrated processes are tied to achievable fuel yields and conversion efficiency of the intermediate steps, i.e., the production of oxygenates/alcohols from syngas and the conversion of oxygenates/alcohols to hydrocarbon fuels.« less

Neutronics Investigations for the Lower Part of a Westinghouse SVEA-96+ Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, M.F.; Luethi, A.; Seiler, R.

2002-05-15

Accurate critical experiments have been performed for the validation of total fission (F{sub tot}) and {sup 238}U-capture (C{sub 8}) reaction rate distributions obtained with CASMO-4, HELIOS, BOXER, and MCNP4B for the lower axial region of a real Westinghouse SVEA-96+ fuel assembly. The assembly comprised fresh fuel with an average {sup 235}U enrichment of 4.02 wt%, a maximum enrichment of 4.74 wt%, 14 burnable-absorber fuel pins, and full-density water moderation. The experimental configuration investigated was core 1A of the LWR-PROTEUS Phase I project, where 61 different fuel pins, representing {approx}64% of the assembly, were gamma-scanned individually. Calculated (C) and measured (E)more » values have been compared in terms of C/E distributions. For F{sub tot}, the standard deviations are 1.2% for HELIOS, 0.9% for CASMO-4, 0.8% for MCNP4B, and 1.7% for BOXER. Standard deviations of 1.1% for HELIOS, CASMO-4, and MCNP4B and 1.2% for BOXER were obtained in the case of C{sub 8}. Despite the high degree of accuracy observed on the average, it was found that the five burnable-absorber fuel pins investigated showed a noticeable underprediction of F{sub tot}, quite systematically, for the deterministic codes evaluated (average C/E for the burnable-absorber fuel pins in the range 0.974 to 0.988, depending on the code)« less

24 CFR 3285.601 - Field assembly.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 24 Housing and Urban Development 5 2011-04-01 2011-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

24 CFR 3285.601 - Field assembly.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 24 Housing and Urban Development 5 2013-04-01 2013-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

24 CFR 3285.601 - Field assembly.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 24 Housing and Urban Development 5 2014-04-01 2014-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

24 CFR 3285.601 - Field assembly.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 24 Housing and Urban Development 5 2012-04-01 2012-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Tianfu; Ma, Zhuang; Li, Guoping

Electrostatic self-assembly in organic solvent without intensively oxidative or corrosive environments, was adopted to prepare Al/Fe{sub 2}O{sub 3}/MWCNT nanostructured energetic materials as an energy generating material. The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles. This spontaneous assembly method without any surfactant chemistry or other chemical and biological moieties decreased the aggregation of the same nanoparticles largely, moreover, the poor interfacial contact between the Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles was improved significantly, which was the key characteristic ofmore » high performance nanostructured energetic materials. In addition, the assembly process was confirmed as Diffusion-Limited Aggregation. The assembled Al/Fe{sub 2}O{sub 3}/MWCNT nanostructured energetic materials showed excellent performance with heat release of 2400 J/g, peak pressure of 0.42 MPa and pressurization rate of 105.71 MPa/s, superior to that in the control group Al/Fe{sub 2}O{sub 3} nanostructured energetic materials prepared by sonication with heat release of 1326 J/g, peak pressure of 0.19 MPa and pressurization rate of 33.33 MPa/s. Therefore, the approach, which is facile, opens a promising route to the high performance nanostructured energetic materials. - Graphical abstract: The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles. - Highlights: • A facile spontaneous electrostatic assembly strategy without surfactant was adopted. • The fuels and oxidizers assembled into densely packed nanostructured composites. • The assembled nanostructured energetic materials have excellent performance. • This high performance energetic material can be scaled up for practical application. • This strategy can be applied into other nanostructured energetic material systems.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Root, M. A.; Menlove, H. O.; Lanza, R. C.

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE PAGES

Root, M. A.; Menlove, H. O.; Lanza, R. C.; ...

2018-03-21

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Analysis of a Neutronic Experiment on a Simulated Mercury Spallation Neutron Target Assembly Bombarded by Giga-Electron-Volt Protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maekawa, Fujio; Meigo, Shin-ichiro; Kasugai, Yoshimi

2005-05-15

A neutronic benchmark experiment on a simulated spallation neutron target assembly was conducted by using the Alternating Gradient Synchrotron at Brookhaven National Laboratory and was analyzed to investigate the prediction capability of Monte Carlo simulation codes used in neutronic designs of spallation neutron sources. The target assembly consisting of a mercury target, a light water moderator, and a lead reflector was bombarded by 1.94-, 12-, and 24-GeV protons, and the fast neutron flux distributions around the target and the spectra of thermal neutrons leaking from the moderator were measured in the experiment. In this study, the Monte Carlo particle transportmore » simulation codes NMTC/JAM, MCNPX, and MCNP-4A with associated cross-section data in JENDL and LA-150 were verified based on benchmark analysis of the experiment. As a result, all the calculations predicted the measured quantities adequately; calculated integral fluxes of fast and thermal neutrons agreed approximately within {+-}40% with the experiments although the overall energy range encompassed more than 12 orders of magnitude. Accordingly, it was concluded that these simulation codes and cross-section data were adequate for neutronics designs of spallation neutron sources.« less

Case study on incentive mechanism of energy efficiency retrofit in coal-fueled power plant in China.

PubMed

Yuan, Donghai; Guo, Xujing; Cao, Yuan; He, Liansheng; Wang, Jinggang; Xi, Beidou; Li, Junqi; Ma, Wenlin; Zhang, Mingshun

2012-01-01

An ordinary steam turbine retrofit project is selected as a case study; through the retrofit, the project activities will generate emission reductions within the power grid for about 92,463 tCO(2)e per annum. The internal rate of return (IRR) of the project is only -0.41% without the revenue of carbon credits, for example, CERs, which is much lower than the benchmark value of 8%. Only when the unit price of carbon credit reaches 125 CNY/tCO(2), the IRR could reach the benchmark and an effective carbon tax needs to increase the price of carbon to 243 CNY/tce in order to make the project financially feasible. Design of incentive mechanism will help these low efficiency enterprises improve efficiency and reduce CO(2) emissions, which can provide the power plants sufficient incentive to implement energy efficiency retrofit project in existing coal-fuel power generation-units, and we hope it will make a good demonstration for the other low efficiency coal-fueled power generation units in China.

Case Study on Incentive Mechanism of Energy Efficiency Retrofit in Coal-Fueled Power Plant in China

PubMed Central

Yuan, Donghai; Guo, Xujing; Cao, Yuan; He, Liansheng; Wang, Jinggang; Xi, Beidou; Li, Junqi; Ma, Wenlin; Zhang, Mingshun

2012-01-01

An ordinary steam turbine retrofit project is selected as a case study; through the retrofit, the project activities will generate emission reductions within the power grid for about 92,463 tCO2e per annum. The internal rate of return (IRR) of the project is only −0.41% without the revenue of carbon credits, for example, CERs, which is much lower than the benchmark value of 8%. Only when the unit price of carbon credit reaches 125 CNY/tCO2, the IRR could reach the benchmark and an effective carbon tax needs to increase the price of carbon to 243 CNY/tce in order to make the project financially feasible. Design of incentive mechanism will help these low efficiency enterprises improve efficiency and reduce CO2 emissions, which can provide the power plants sufficient incentive to implement energy efficiency retrofit project in existing coal-fuel power generation-units, and we hope it will make a good demonstration for the other low efficiency coal-fueled power generation units in China. PMID:23365532

Molybdenum-99 production calculation analysis of SAMOP reactor based on thorium nitrate fuel

NASA Astrophysics Data System (ADS)

Syarip; Togatorop, E.; Yassar

2018-03-01

SAMOP (Subcritical Assembly for Molybdenum-99 Production) has the potential to use thorium as fuel to produce 99Mo after modifying the design, but the production performance has not been discovered yet. A study needs to be done to obtain the correlation between 99Mo production with the mixed fuel composition of uranium and with SAMOP power on the modified SAMOP design. The study aims to obtain the production of 99Mo based thorium nitrate fuel on SAMOP’s modified designs. Monte Carlo N-Particle eXtended (MCNPX) is required to simulate the operation of the assembly by varying the composition of the uranium-thorium nitrate mixed fuel, geometry and power fraction on the SAMOP modified designs. The burnup command on the MCNPX is used to confirm the 99Mo production result. The assembly is simulated to operate for 6 days with subcritical neutron multiplication factor (keff = 0.97-0.99). The neutron multiplication factor of the modified design (keff) is 0.97, the activity obtained from 99Mo is 18.58 Ci at 1 kW power operation.

Premixer assembly for mixing air and fuel for combustion

DOEpatents

York, William David; Johnson, Thomas Edward; Keener, Christopher Paul

2016-12-13

A premixer assembly for mixing air and fuel for combustion includes a plurality of tubes disposed at a head end of a combustor assembly. Also included is a tube of the plurality of tubes, the tube including an inlet end and an outlet end. Further included is at least one non-circular portion of the tube extending along a length of the tube, the at least one non-circular portion having a non-circular cross-section, and the tube having a substantially constant cross-sectional area along its length

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M. M.

1998-10-14

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required tomore » match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.« less

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lubina, A. S., E-mail: lubina-as@nrcki.ru; Subbotin, A. S.; Sedov, A. A.

2016-12-15

The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The resultsmore » of the hydrodynamics and heat transfer calculations have been analyzed.« less

Biomass fuels update. TVAs biomass fuels program

NASA Astrophysics Data System (ADS)

1982-02-01

Equipment was installed and tests were conducted on the ethanol from hardwood project. Location of hardwoods, to improve forest management, and to reduce the cost of harvesting woody biomass was assessed. Substantial underutilized cropland exists in the Valley, and a questionnaire survey was administered to supplement available cropland data. The potential liquid fuel yields and production management practices for alternative starch, sugar, and vegetable oil crops were determined to obtain benchmark data and to evaluate alcohol production from alternative agricultural feedstocks. Workshops were conducted to provide information on production of alcohol.

HSPES membrane electrode assembly

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor); Yen, Shiao-Ping (Inventor)

2000-01-01

An improved fuel cell electrode, as well as fuel cells and membrane electrode assemblies that include such an electrode, in which the electrode includes a backing layer having a sintered layer thereon, and a non-sintered free-catalyst layer. The invention also features a method of forming the electrode by sintering a backing material with a catalyst material and then applying a free-catalyst layer.

5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND ...

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

5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND LOWER FUEL ELEMENTS. TAKEN FROM TOP OF SHIELDING TANK WITH CAMERA POINTING TOWARDS BOTTOM OF TANK. SHOWS LADDER, SQUARE LIFTING FRAME, FUEL ELEMENT HOLDERS, AND CABLE CYLINDERS. INEL PHOTO NUMBER 65-5434, TAKEN OCTOBER 20, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pytel, K.; Mieleszczenko, W.; Lechniak, J.

2010-03-01

The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

Methods and systems to thermally protect fuel nozzles in combustion systems

DOEpatents

Helmick, David Andrew; Johnson, Thomas Edward; York, William David; Lacy, Benjamin Paul

2013-12-17

A method of assembling a gas turbine engine is provided. The method includes coupling a combustor in flow communication with a compressor such that the combustor receives at least some of the air discharged by the compressor. A fuel nozzle assembly is coupled to the combustor and includes at least one fuel nozzle that includes a plurality of interior surfaces, wherein a thermal barrier coating is applied across at least one of the plurality of interior surfaces to facilitate shielding the interior surfaces from combustion gases.

Green-fuel-mediated synthesis of self-assembled NiO nano-sticks for dual applications—photocatalytic activity on Rose Bengal dye and antimicrobial action on bacterial strains

NASA Astrophysics Data System (ADS)

Iyyappa Rajan, P.; Vijaya, J. Judith; Jesudoss, S. K.; Kaviyarasu, K.; Kennedy, L. John; Jothiramalingam, R.; Al-Lohedan, Hamad A.; Vaali-Mohammed, Mansoor-Ali

2017-08-01

With aim of promoting the employability of green fuels in the synthesis of nano-scaled materials with new kinds of morphologies for multiple applications, successful synthesis of self-assembled NiO nano-sticks was achieved through a 100% green-fuel-mediated hot-plate combustion reaction. The synthesized NiO nano-sticks show excellent photocatalytic activity on Rose Bengal dye and superior antibacterial potential towards both Gram-positive and Gram-negative bacteria.

Advanced membrane electrode assemblies for fuel cells

DOEpatents

Kim, Yu Seung; Pivovar, Bryan S.

2012-07-24

A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

General view of a Space Shuttle Main Engine (SSME) mounted ...

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

General view of a Space Shuttle Main Engine (SSME) mounted on an SSME engine handler, taken in the SSME Processing Facility at Kennedy Space Center. The most prominent features of the engine assembly in this view are the Low-Pressure Fuel Turbopump Discharge Duct looping diagonally across the top of the assembly and connecting to the High-Pressure Fuel Turbopump, the Low-Pressure Oxidizer Turbopump (LPOTP) located center right of the assembly and the LPOTP Discharge Duct looping around from the pump to the underside of the engine assembly in this view. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

Advanced membrane electrode assemblies for fuel cells

DOEpatents

Kim, Yu Seung; Pivovar, Bryan S

2014-02-25

A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

NASA Astrophysics Data System (ADS)

Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

2015-07-01

During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

Neutronic fuel element fabrication

DOEpatents

Korton, George

2004-02-24

This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.

Sustained Recycle in Light Water and Sodium-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2010-11-01

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

Best Practices and Testing Protocols for Benchmarking ORR Activities of Fuel Cell Electrocatalysts Using Rotating Disk Electrode

DOE PAGES

Kocha, Shyam S.; Shinozaki, Kazuma; Zack, Jason W.; ...

2017-05-02

Thin-film-rotating disk electrodes (TF-RDEs) are the half-cell electrochemical system of choice for rapid screening of oxygen reduction reaction (ORR) activity of novel Pt supported on carbon black supports (Pt/C) electrocatalysts. It has been shown that the magnitude of the measured ORR activity and reproducibility are highly dependent on the system cleanliness, evaluation protocols, and operating conditions as well as ink formulation, composition, film drying, and the resultant film thickness and uniformity. Accurate benchmarks of baseline Pt/C catalysts evaluated using standardized protocols and best practices are necessary to expedite ultra-low-platinum group metal (PGM) catalyst development that is crucial for the imminentmore » commercialization of fuel cell vehicles. We report results of evaluation in three independent laboratories of Pt/C electrocatalysts provided by commercial fuel cell catalyst manufacturers (Johnson Matthey, Umicore, Tanaka Kikinzoku Kogyo - TKK). The studies were conducted using identical evaluation protocols/ink formulation/film fabrication albeit employing unique electrochemical cell designs specific to each laboratory. Furthermore, the ORR activities reported in this work provide a baseline and criteria for selection and scale-up of novel high activity ORR electrocatalysts for implementation in proton exchange membrane fuel cells (PEMFCs).« less

Reconciling sectoral abatement strategies with global climate targets: the case of the Chinese passenger vehicle fleet.

PubMed

Pauliuk, Stefan; Dhaniati, Ni Made A; Müller, Daniel B

2012-01-03

The IPCC Forth Assessment Report postulates that global warming can be limited to 2 °C by deploying technologies that are currently available or expected to be commercialized in the coming decades. However, neither specific technological pathways nor internationally binding reduction targets for different sectors or countries have been established yet. Using the passenger car stock in China as example we compute direct CO(2) emissions until 2050 depending on population, car utilization, and fuel efficiency and compare them to benchmarks derived by assuming even contribution of all sectors and a unitary global per capita emission quota. Compared to present car utilization in industrialized countries, massive deployment of prototypes of fuel efficient cars could reduce emissions by about 45%, and moderately lower car use could contribute with another 33%. Still, emissions remain about five times higher than the benchmark for the 2 °C global warming target. Therefore an extended analysis, including in particular low-carbon fuels and the impact of urban and transport planning on annual distance traveled and car ownership, should be considered. A cross-sectoral comparison could reveal whether other sectors could bear an overproportional reduction quota instead. The proposed model offers direct interfaces to material industries, fuel production, and scrap vehicle supply.

Potential for Electrified Vehicles to Contribute to U.S. Petroleum and Climate Goals and Implications for Advanced Biofuels.

PubMed

Meier, Paul J; Cronin, Keith R; Frost, Ethan A; Runge, Troy M; Dale, Bruce E; Reinemann, Douglas J; Detlor, Jennifer

2015-07-21

To examine the national fuel and emissions impacts from increasingly electrified light-duty transportation, we reconstructed the vehicle technology portfolios from two national vehicle studies. Using these vehicle portfolios, we normalized assumptions and examined sensitivity around the rates of electrified vehicle penetration, travel demand growth, and electricity decarbonization. We further examined the impact of substituting low-carbon advanced cellulosic biofuels in place of petroleum. Twenty-seven scenarios were benchmarked against a 50% petroleum-reduction target and an 80% GHG-reduction target. We found that with high rates of electrification (40% of miles traveled) the petroleum-reduction benchmark could be satisfied, even with high travel demand growth. The same highly electrified scenarios, however, could not satisfy 80% GHG-reduction targets, even assuming 80% decarbonized electricity and no growth in travel demand. Regardless of precise consumer vehicle preferences, emissions are a function of the total reliance on electricity versus liquid fuels and the corresponding greenhouse gas intensities of both. We found that at a relatively high rate of electrification (40% of miles and 26% by fuel), an 80% GHG reduction could only be achieved with significant quantities of low-carbon liquid fuel in cases with low or moderate travel demand growth.

Best Practices and Testing Protocols for Benchmarking ORR Activities of Fuel Cell Electrocatalysts Using Rotating Disk Electrode

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kocha, Shyam S.; Shinozaki, Kazuma; Zack, Jason W.

Abstract Thin-film-rotating disk electrodes (TF-RDEs) are the half-cell electrochemical system of choice for rapid screening of oxygen reduction reaction (ORR) activity of novel Pt supported on carbon black supports (Pt/C) electrocatalysts. It has been shown that the magnitude of the measured ORR activity and reproducibility are highly dependent on the system cleanliness, evaluation protocols, and operating conditions as well as ink formulation, composition, film drying, and the resultant film thickness and uniformity. Accurate benchmarks of baseline Pt/C catalysts evaluated using standardized protocols and best practices are necessary to expedite ultra-low-platinum group metal (PGM) catalyst development that is crucial for themore » imminent commercialization of fuel cell vehicles. We report results of evaluation in three independent laboratories of Pt/C electrocatalysts provided by commercial fuel cell catalyst manufacturers (Johnson Matthey, Umicore, Tanaka Kikinzoku Kogyo—TKK). The studies were conducted using identical evaluation protocols/ink formulation/film fabrication albeit employing unique electrochemical cell designs specific to each laboratory. The ORR activities reported in this work provide a baseline and criteria for selection and scale-up of novel high activity ORR electrocatalysts for implementation in proton exchange membrane fuel cells (PEMFCs).« less

Best Practices and Testing Protocols for Benchmarking ORR Activities of Fuel Cell Electrocatalysts Using Rotating Disk Electrode

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kocha, Shyam S.; Shinozaki, Kazuma; Zack, Jason W.

Thin-film-rotating disk electrodes (TF-RDEs) are the half-cell electrochemical system of choice for rapid screening of oxygen reduction reaction (ORR) activity of novel Pt supported on carbon black supports (Pt/C) electrocatalysts. It has been shown that the magnitude of the measured ORR activity and reproducibility are highly dependent on the system cleanliness, evaluation protocols, and operating conditions as well as ink formulation, composition, film drying, and the resultant film thickness and uniformity. Accurate benchmarks of baseline Pt/C catalysts evaluated using standardized protocols and best practices are necessary to expedite ultra-low-platinum group metal (PGM) catalyst development that is crucial for the imminentmore » commercialization of fuel cell vehicles. We report results of evaluation in three independent laboratories of Pt/C electrocatalysts provided by commercial fuel cell catalyst manufacturers (Johnson Matthey, Umicore, Tanaka Kikinzoku Kogyo - TKK). The studies were conducted using identical evaluation protocols/ink formulation/film fabrication albeit employing unique electrochemical cell designs specific to each laboratory. Furthermore, the ORR activities reported in this work provide a baseline and criteria for selection and scale-up of novel high activity ORR electrocatalysts for implementation in proton exchange membrane fuel cells (PEMFCs).« less

Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system

DOEpatents

Christiansen, David W.; Smith, Bob G.

1982-01-01

A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lafleur, Adrienne M.; Ulrich, Timothy J. II; Menlove, Howard O.

Objective is to investigate the use of Passive Neutron Albedo Reactivity (PNAR) and Self-Interrogation Neutron Resonance Densitometry (SINRD) to quantify fissile content in FUGEN spent fuel assemblies (FAs). Methodology used is: (1) Detector was designed using fission chambers (FCs); (2) Optimized design via MCNPX simulations; and (3) Plan to build and field test instrument in FY13. Significance was to improve safeguards verification of spent fuel assemblies in water and increase sensitivity to partial defects. MCNPX simulations were performed to optimize the design of the SINRD+PNAR detector. PNAR ratio was less sensitive to FA positioning than SINRD and SINRD ratio wasmore » more sensitive to Pu fissile mass than PNAR. Significance was that the integration of these techniques can be used to improve verification of spent fuel assemblies in water.« less

Next Generation Safeguards Initiative research to determine the Pu mass in spent fuel assemblies: Purpose, approach, constraints, implementation, and calibration

NASA Astrophysics Data System (ADS)

Tobin, S. J.; Menlove, H. O.; Swinhoe, M. T.; Schear, M. A.

2011-10-01

The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy has funded a multi-lab/multi-university collaboration to quantify the plutonium mass in spent nuclear fuel assemblies and to detect the diversion of pins from them. The goal of this research effort is to quantify the capability of various non-destructive assay (NDA) technologies as well as to train a future generation of safeguards practitioners. This research is "technology driven" in the sense that we will quantify the capabilities of a wide range of safeguards technologies of interest to regulators and policy makers; a key benefit to this approach is that the techniques are being tested in a unified manner. When the results of the Monte Carlo modeling are evaluated and integrated, practical constraints are part of defining the potential context in which a given technology might be applied. This paper organizes the commercial spent fuel safeguard needs into four facility types in order to identify any constraints on the NDA system design. These four facility types are the following: future reprocessing plants, current reprocessing plants, once-through spent fuel repositories, and any other sites that store individual spent fuel assemblies (reactor sites are the most common facility type in this category). Dry storage is not of interest since individual assemblies are not accessible. This paper will overview the purpose and approach of the NGSI spent fuel effort and describe the constraints inherent in commercial fuel facilities. It will conclude by discussing implementation and calibration of measurement systems. This report will also provide some motivation for considering a couple of other safeguards concepts (base measurement and fingerprinting) that might meet the safeguards need but not require the determination of plutonium mass.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE PAGES

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

2017-03-27

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Materials and Fuels Complex Tour

ScienceCinema

Miley, Don

2017-12-11

The Materials and Fuels Complex at Idaho National Laboratory is home to several facilities used for the research and development of nuclear fuels. Stops include the Fuel Conditioning Facility, the Hot Fuel Examination Facility (post-irradiation examination), and the Space and Security Power System Facility, where radioisotope thermoelectric generators (RTGs) are assembled for deep space missions.

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2014 CFR

2014-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2012 CFR

2012-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2011 CFR

2011-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2013 CFR

2013-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

30 CFR 36.27 - Fuel-supply system.

Code of Federal Regulations, 2010 CFR

2010-07-01

... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

QUAST: quality assessment tool for genome assemblies

PubMed Central

Gurevich, Alexey; Saveliev, Vladislav; Vyahhi, Nikolay; Tesler, Glenn

2013-01-01

Summary: Limitations of genome sequencing techniques have led to dozens of assembly algorithms, none of which is perfect. A number of methods for comparing assemblers have been developed, but none is yet a recognized benchmark. Further, most existing methods for comparing assemblies are only applicable to new assemblies of finished genomes; the problem of evaluating assemblies of previously unsequenced species has not been adequately considered. Here, we present QUAST—a quality assessment tool for evaluating and comparing genome assemblies. This tool improves on leading assembly comparison software with new ideas and quality metrics. QUAST can evaluate assemblies both with a reference genome, as well as without a reference. QUAST produces many reports, summary tables and plots to help scientists in their research and in their publications. In this study, we used QUAST to compare several genome assemblers on three datasets. QUAST tables and plots for all of them are available in the Supplementary Material, and interactive versions of these reports are on the QUAST website. Availability: http://bioinf.spbau.ru/quast Contact: gurevich@bioinf.spbau.ru Supplementary information: Supplementary data are available at Bioinformatics online. PMID:23422339

Inlet nozzle assembly

DOEpatents

Christiansen, David W.; Karnesky, Richard A.; Precechtel, Donald R.; Smith, Bob G.; Knight, Ronald C.

1987-01-01

An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

Inlet nozzle assembly

DOEpatents

Christiansen, D.W.; Karnesky, R.A.; Knight, R.C.; Precechtel, D.R.; Smith, B.G.

1985-09-09

An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn Thomas; De Baere, Paul

2015-04-17

A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hourmore » of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.« less

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

DOE PAGES

Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.; ...

2015-10-09

Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.

Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

Open Rotor - Analysis of Diagnostic Data

NASA Technical Reports Server (NTRS)

Envia, Edmane

2011-01-01

NASA is researching open rotor propulsion as part of its technology research and development plan for addressing the subsonic transport aircraft noise, emission and fuel burn goals. The low-speed wind tunnel test for investigating the aerodynamic and acoustic performance of a benchmark blade set at the approach and takeoff conditions has recently concluded. A high-speed wind tunnel diagnostic test campaign has begun to investigate the performance of this benchmark open rotor blade set at the cruise condition. Databases from both speed regimes will comprise a comprehensive collection of benchmark open rotor data for use in assessing/validating aerodynamic and noise prediction tools (component & system level) as well as providing insights into the physics of open rotors to help guide the development of quieter open rotors.

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

DOEpatents

Stengel, F.G.

1963-12-24

A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

High performance internal reforming unit for high temperature fuel cells

DOEpatents

Ma, Zhiwen [Sandy Hook, CT; Venkataraman, Ramakrishnan [New Milford, CT; Novacco, Lawrence J [Brookfield, CT

2008-10-07

A fuel reformer having an enclosure with first and second opposing surfaces, a sidewall connecting the first and second opposing surfaces and an inlet port and an outlet port in the sidewall. A plate assembly supporting a catalyst and baffles are also disposed in the enclosure. A main baffle extends into the enclosure from a point of the sidewall between the inlet and outlet ports. The main baffle cooperates with the enclosure and the plate assembly to establish a path for the flow of fuel gas through the reformer from the inlet port to the outlet port. At least a first directing baffle extends in the enclosure from one of the sidewall and the main baffle and cooperates with the plate assembly and the enclosure to alter the gas flow path. Desired graded catalyst loading pattern has been defined for optimized thermal management for the internal reforming high temperature fuel cells so as to achieve high cell performance.

Road user fee task force report to the 72nd Oregon Legislative Assembly on the possible alternatives of the current system of taxing highway use through motor vehicle fuel taxes

DOT National Transportation Integrated Search

2003-03-01

Recognizing that the fuel tax is a declining revenue source for Oregon's road system, the 2001 Oregon Legislative Assembly passed House Bill 3946, mandating the formation of the Road User Fee Task Force with the mission to develop a design for revenu...

Manifold seal for fuel cell stack assembly

DOEpatents

Schmitten, Phillip F.; Wright, Maynard K.

1989-01-01

An assembly for sealing a manifold to a stack of fuel cells includes a first resilient member for providing a first sealing barrier between the manifold and the stack. A second resilient member provides a second sealing barrier between the manifold and the stack. The first and second resilient members are retained in such a manner as to define an area therebetween adapted for retaining a sealing composition.

Fuel assembly shaker and truck test simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klymyshyn, Nicholas A.; Jensen, Philip J.; Sanborn, Scott E.

2014-09-30

This study continues the modeling support of the SNL shaker table task from 2013 and includes analysis of the SNL 2014 truck test campaign. Detailed finite element models of the fuel assembly surrogate used by SNL during testing form the basis of the modeling effort. Additional analysis was performed to characterize and filter the accelerometer data collected during the SNL testing. The detailed fuel assembly finite element model was modified to improve the performance and accuracy of the original surrogate fuel assembly model in an attempt to achieve a closer agreement with the low strains measured during testing. The revisedmore » model was used to recalculate the shaker table load response from the 2013 test campaign. As it happened, the results remained comparable to the values calculated with the original fuel assembly model. From this it is concluded that the original model was suitable for the task and the improvements to the model were not able to bring the calculated strain values down to the extremely low level recorded during testing. The model needs more precision to calculate strains that are so close to zero. The truck test load case had an even lower magnitude than the shaker table case. Strain gage data from the test was compared directly to locations on the model. Truck test strains were lower than the shaker table case, but the model achieved a better relative agreement of 100-200 microstrains (or 0.0001-0.0002 mm/mm). The truck test data included a number of accelerometers at various locations on the truck bed, surrogate basket, and surrogate fuel assembly. This set of accelerometers allowed an evaluation of the dynamics of the conveyance system used in testing. It was discovered that the dynamic load transference through the conveyance has a strong frequency-range dependency. This suggests that different conveyance configurations could behave differently and transmit different magnitudes of loads to the fuel even when traveling down the same road at the same speed. It is recommended that the SNL conveyance system used in testing be characterized through modal analysis and frequency response analysis to provide context and assist in the interpretation of the strain data that was collected during the truck test campaign.« less

Measurement of the 235U Induced Fission Gamma-ray Spectrum as an Active Non-destructive Assay of Fresh Nucleear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sarnoski, Sarah E.; Fast, James E.; Fulsom, Bryan G.

2017-07-17

Non-destructive assay is a powerful tool the International Atomic Energy Agency (IAEA) employs to verify adherence to safeguards agreements. Current IAEA veri- cation techniques for fresh nuclear fuel include passive gamma-ray spectroscopy to determine fuel enrichment. This technique suers from self-shielding and lakes the percision to detect diversion of central fuel rods. The aim of this research is to develop a new, more capable non-destructive analysis technique using active neutron interroga- tion of fuel assemblies and determining the yields of short-lived ssion products from high-resolution gamma-ray spectroscopy using high-purity germanium (HPGe). This paper reports results from irradiation of a onemore » meter tall mock fresh fuel assembly with low enriched uranium (LEU) or depleted uranium (DU) rods using a down-scattered deuterium-tritium (D-T) neutron source. Both prompt and delayed gamma-ray spec- tra were collected as time-stamped list-mode data in a coax detector and without list mode data in a planar strip detector. No dierentiating signatures were observed in the prompt spectra in either detector; however, both detectors observed several short-lived ssion product signatures in LEU and not DU fuel, indicating that this technique has potential for determination of enrichment of fresh fuel assemblies. There were eight unique ssion products observed in the LEU spectra with the coax detector spectra, and three ssion products were observed in the LEU spectra with the strip detector.« less

Downhole steam generator with improved preheating/cooling features

DOEpatents

Donaldson, A. Burl; Hoke, Donald E.; Mulac, Anthony J.

1983-01-01

An apparatus for downhole steam generation employing dual-stage preheaters for liquid fuel and for the water. A first heat exchange jacket for the fuel surrounds the fuel/oxidant mixing section of the combustor assembly downstream of the fuel nozzle and contacts the top of the combustor unit of the combustor assembly, thereby receiving heat directly from the combustion of the fuel/oxidant. A second stage heat exchange jacket surrounds an upper portion of the oxidant supply line adjacent the fuel nozzle receiving further heat from the compression heat which results from pressurization of the oxidant. The combustor unit includes an inner combustor sleeve whose inner wall defines the combustion zone. The inner combustor sleeve is surrounded by two concentric water channels, one defined by the space between the inner combustor sleeve and an intermediate sleeve, and the second defined by the space between the intermediate sleeve and an outer cylindrical housing. The channels are connected by an annular passage adjacent the top of the combustor assembly and the countercurrent nature of the water flow provides efficient cooling of the inner combustor sleeve. An annular water ejector with a plurality of nozzles is provided to direct water downwardly into the combustor unit at the boundary of the combustion zone and along the lower section of the intermediate sleeve.

Downhole steam generator with improved preheating/cooling features. [Patent application

DOEpatents

Donaldson, A.B.; Hoke, D.E.; Mulac, A.J.

1980-10-10

An apparatus is described for downhole steam generation employing dual-stage preheaters for liquid fuel and for the water. A first heat exchange jacket for the fuel surrounds the fuel/oxidant mixing section of the combustor assembly downstream of the fuel nozzle and contacts the top of the combustor unit of the combustor assembly, thereby receiving heat directly from the combustion of the fuel/oxidant. A second stage heat exchange jacket surrounds an upper portion of the oxidant supply line adjacent the fuel nozzle receiving further heat from the compression heat which results from pressurization of the oxidant. The combustor unit includes an inner combustor sleeve whose inner wall defines the combustion zone. The inner combustor sleeve is surrounded by two concentric water channels, one defined by the space between the inner combustor sleeve and an intermediate sleeve, and the second defined by the space between the intermediate sleeve and an outer cylindrical housing. The channels are connected by an annular passage adjacent the top of the combustor assembly and the countercurrent nature of the water flow provides efficient cooling of the inner combustor sleeve. An annular water ejector with a plurality of nozzles is provided to direct water downwardly into the combustor unit at the boundary of the combustion zone and along the lower section of the intermediate sleeve.

High energy X-ray CT study on the central void formations and the fuel pin deformations of FBR fuel assemblies

NASA Astrophysics Data System (ADS)

Katsuyama, Kozo; Nagamine, Tsuyoshi; Matsumoto, Shin-ichiro; Sato, Seichi

2007-02-01

The central void formations and deformations of fuel pins were investigated in fuel assemblies irradiated to high burn-up, using a non-destructive X-ray CT (computer tomography) technique. In this X-ray CT, the effect of strong gamma ray activity could be reduced to a negligible degree by using the pulse of a high energy X-ray source and detecting the intensity of the transmitted X-rays in synchronization with the generated X-rays. Clear cross-sectional images of fuel assemblies irradiated to high burn-up in a fast breeder reactor were successively obtained, in which the wrapping wires, cladding, pellets and central voids could be distinctly seen. The diameter of a typical central void measured by X-ray CT agreed with the one obtained by ceramography within an error of 0.1 mm. Based on this result, the dependence of the central void diameter on the linear heating rate was analyzed. In addition, the deformation behavior of a fuel pin along its axial direction could be analyzed from 20 stepwise X-ray cross-sectional images obtained in a small interval, and the results obtained showed a good agreement with the predictions calculated by two computer codes.

Fuel cell with metal screen flow-field

DOEpatents

Wilson, M.S.; Zawodzinski, C.

1998-08-25

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field there between for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells. 11 figs.

Fuel cell with metal screen flow-field

DOEpatents

Wilson, Mahlon S.; Zawodzinski, Christine

2001-01-01

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field therebetween for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells.

Fuel cell with metal screen flow-field

DOEpatents

Wilson, Mahlon S.; Zawodzinski, Christine

1998-01-01

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field therebetween for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells.

Generator module architecture for a large solid oxide fuel cell power plant

DOEpatents

Gillett, James E.; Zafred, Paolo R.; Riggle, Matthew W.; Litzinger, Kevin P.

2013-06-11

A solid oxide fuel cell module contains a plurality of integral bundle assemblies, the module containing a top portion with an inlet fuel plenum and a bottom portion receiving air inlet feed and containing a base support, the base supports dense, ceramic exhaust manifolds which are below and connect to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the fuel cells comprise a fuel cell stack bundle all surrounded within an outer module enclosure having top power leads to provide electrical output from the stack bundle, where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all 100% of the weight of the stack, and each bundle assembly has its own control for vertical and horizontal thermal expansion control.

Fuel cell assembly unit for promoting fluid service and electrical conductivity

DOEpatents

Jones, Daniel O.

1999-01-01

Fluid service and/or electrical conductivity for a fuel cell assembly is promoted. Open-faced flow channel(s) are formed in a flow field plate face, and extend in the flow field plate face between entry and exit fluid manifolds. A resilient gas diffusion layer is located between the flow field plate face and a membrane electrode assembly, fluidly serviced with the open-faced flow channel(s). The resilient gas diffusion layer is restrained against entering the open-faced flow channel(s) under a compressive force applied to the fuel cell assembly. In particular, a first side of a support member abuts the flow field plate face, and a second side of the support member abuts the resilient gas diffusion layer. The support member is formed with a plurality of openings extending between the first and second sides of the support member. In addition, a clamping pressure is maintained for an interface between the resilient gas diffusion layer and a portion of the membrane electrode assembly. Preferably, the support member is spikeless and/or substantially flat. Further, the support member is formed with an electrical path for conducting current between the resilient gas diffusion layer and position(s) on the flow field plate face.

CASMO5/TSUNAMI-3D spent nuclear fuel reactivity uncertainty analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferrer, R.; Rhodes, J.; Smith, K.

2012-07-01

The CASMO5 lattice physics code is used in conjunction with the TSUNAMI-3D sequence in ORNL's SCALE 6 code system to estimate the uncertainties in hot-to-cold reactivity changes due to cross-section uncertainty for PWR assemblies at various burnup points. The goal of the analysis is to establish the multiplication factor uncertainty similarity between various fuel assemblies at different conditions in a quantifiable manner and to obtain a bound on the hot-to-cold reactivity uncertainty over the various assembly types and burnup attributed to fundamental cross-section data uncertainty. (authors)

Method for recovering catalytic elements from fuel cell membrane electrode assemblies

DOEpatents

Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ; Heinz, Robert [Ludwigshafen, DE

2012-06-26

A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

Fail-safe storage rack for irradiated fuel rod assemblies

DOEpatents

Lewis, D.R.

1993-03-23

A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

Fail-safe storage rack for irradiated fuel rod assemblies

DOEpatents

Lewis, Donald R.

1993-01-01

A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

Management of thermal peaking factors in CONFU-B PWR assemblies using neutron poisons and tailored enrichment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Visosky, M.; Hejzlar, P.; Kazimi, M.

2006-07-01

CONFU-B assemblies are PWR assemblies containing standard Uranium fuel rods and TRU bearing inert material fuel rods and are designed to achieve net TRU destruction over a 4.5-year irradiation. These highly heterogeneous assemblies tend to exhibit large intra-assembly power peaking factors (IAPPF). Neutronic strategies to reduce IAPPF are developed. The IAPPF are calculated at the assembly level using CASMO4, and these are used to calculate the most restrictive thermal margin (the Minimum Departure from Nucleate Boiling Ratio, MDNBR) using a whole-core VIPRE-01 model. This paper examines two strategies to manage the thermal margin of a CONFU-B assembly while retaining themore » TRU destruction performance: use of neutron poisons and tailored enrichment schemes. Burnable poisons can be used to suppress BOL reactivity of fresh CONFU-B assemblies with only minor impact on MDNBR and TRU destruction performance. Tailored enrichment, along with the use of soluble boron, can achieve significant improvements in MDNBR, but at some cost to TRU destruction performance. (authors)« less

Optical fuel pin scanner. [Patent application; for reading identifications

DOEpatents

Kirchner, T.L.; Powers, H.G.

1980-12-09

This patent relates to an optical identification system developed for post-irradiation disassembly and analysis of fuel bundle assemblies. The apparatus is designed to be lowered onto a stationary fuel pin to read identification numbers or letters imprinted on the circumference of the top fuel pin and cap. (DLC)

SPACE PROPULSION SYSTEM PHASED-MISSION PROBABILITY ANALYSIS USING CONVENTIONAL PRA METHODS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curtis Smith; James Knudsen

As part of a series of papers on the topic of advance probabilistic methods, a benchmark phased-mission problem has been suggested. This problem consists of modeling a space mission using an ion propulsion system, where the mission consists of seven mission phases. The mission requires that the propulsion operate for several phases, where the configuration changes as a function of phase. The ion propulsion system itself consists of five thruster assemblies and a single propellant supply, where each thruster assembly has one propulsion power unit and two ion engines. In this paper, we evaluate the probability of mission failure usingmore » the conventional methodology of event tree/fault tree analysis. The event tree and fault trees are developed and analyzed using Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE). While the benchmark problem is nominally a "dynamic" problem, in our analysis the mission phases are modeled in a single event tree to show the progression from one phase to the next. The propulsion system is modeled in fault trees to account for the operation; or in this case, the failure of the system. Specifically, the propulsion system is decomposed into each of the five thruster assemblies and fed into the appropriate N-out-of-M gate to evaluate mission failure. A separate fault tree for the propulsion system is developed to account for the different success criteria of each mission phase. Common-cause failure modeling is treated using traditional (i.e., parametrically) methods. As part of this paper, we discuss the overall results in addition to the positive and negative aspects of modeling dynamic situations with non-dynamic modeling techniques. One insight from the use of this conventional method for analyzing the benchmark problem is that it requires significant manual manipulation to the fault trees and how they are linked into the event tree. The conventional method also requires editing the resultant cut sets to obtain the correct results. While conventional methods may be used to evaluate a dynamic system like that in the benchmark, the level of effort required may preclude its use on real-world problems.« less

Nuclear fuel element nut retainer cup. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walton, L.A.

1977-07-19

A typical embodiment has an end fitting for a nuclear reactor fuel element that is joined to the control rod guide tubes by means of a nut plate assembly. The nut plate assembly has an array of nuts, each engaging the respective threaded end of the control rod guide tubes. The nuts, moreover, are retained on the plate during handling and before fuel element assembly by means of hollow cylindrical locking cups that are brazed to the plate and loosely circumscribe the individual enclosed nuts. After the nuts are threaded onto the respective guide tube ends, the locking cups aremore » partially deformed to prevent one or more of the nuts from working loose during reactor operation. The locking cups also prevent loose or broken end fitting parts from becoming entrained in the reactor coolant.« less

Systems and methods for preventing flashback in a combustor assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, Thomas Edward; Ziminsky, Willy Steve; Stevenson, Christian Xavier

2016-04-05

Embodiments of the present application include a combustor assembly. The combustor assembly may include a combustion chamber, a first plenum, a second plenum, and one or more elongate air/fuel premixing injection tubes. Each of the elongate air/fuel premixing injection tubes may include a first length at least partially disposed within the first plenum and configured to receive a first fluid from the first plenum. Moreover, each of the elongate air/fuel premixing injection tubes may include a second length disposed downstream of the first length and at least partially disposed within the second plenum. The second length may be formed ofmore » a porous wall configured to allow a second fluid from the second plenum to enter the second length and create a boundary layer about the porous wall.« less

Membrane electrode gasket assembly (MEGA) technology for polymer electrolyte fuel cells

NASA Astrophysics Data System (ADS)

Pozio, A.; Giorgi, L.; De Francesco, M.; Silva, R. F.; Lo Presti, R.; Danzi, A.

A new technology for the production of a membrane electrode gasket assembly (MEGA) for polymer electrolyte fuel cells (PEFCs) is defined. The MEGA system was prepared by sealing a previously prepared membrane electrode assembly (MEA) in a moulded gasket. For this aim, a proprietary silicone based liquid mixture was injected directly into the MEA borders. Gaskets obtained in different shapes and hardness grades are stable in a wide temperature range. The MEGA technology shows several advantages with respect to traditional PEFCs stack assembling systems: effective membrane saving, reduced fabrication time, possibility of quality control and failed elements substitution. This technology was successfully tested at the ENEA laboratories and the results were acquired in laboratory scale, but industrial production appears to be simple and cheap.

Integral Fast Reactor fuel pin processor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Levinskas, D.

1993-01-01

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

Integral Fast Reactor fuel pin processor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Levinskas, D.

1993-03-01

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

Thermal analysis of the FSP-1 fuel pin irradiation test. [for SP-100 space power reactor

NASA Technical Reports Server (NTRS)

Lyon, William F., III

1991-01-01

Thermal analysis of a pin from the FSP-1 fuels irradiation test has been completed. The purpose of the analysis was to provide predictions of fuel pin temperatures, determine the flow regime within the lithium annulus of the test assembly, and provide a standardized model for a consistent basis of comparison between pins within the test assembly. The calculations have predicted that the pin is operating at slightly above the test design temperatures and that the flow regime within the lithium annulus is a laminar buoyancy driven flow.

In situ observations of water production and distribution in an operating H2/O2 PEM fuel cell assembly using 1H NMR microscopy.

PubMed

Feindel, Kirk W; LaRocque, Logan P-A; Starke, Dieter; Bergens, Steven H; Wasylishen, Roderick E

2004-09-22

Proton NMR imaging was used to investigate in situ the distribution of water in a polymer electrolyte membrane fuel cell operating on H2 and O2. In a single experiment, water was monitored in the gas flow channels, the membrane electrode assembly, and in the membrane surrounding the catalysts. Radial gradient diffusion removes water from the catalysts into the surrounding membrane. This research demonstrates the strength of 1H NMR microscopy as an aid for designing fuel cells to optimize water management.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanan, N. A.; Matos, J. E.

At The request of the Czech Technical University in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. For core configurations C1 and C2, criticality calculations were done for cases with all control rodsmore » at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were done for the C1 core configuration. Finally the reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations.« less

Dry compliant seal for phosphoric acid fuel cell

DOEpatents

Granata, Jr., Samuel J.; Woodle, Boyd M.

1990-01-01

A dry compliant overlapping seal for a phosphoric acid fuel cell preformed f non-compliant Teflon to make an anode seal frame that encircles an anode assembly, a cathode seal frame that encircles a cathode assembly and a compliant seal frame made of expanded Teflon, generally encircling a matrix assembly. Each frame has a thickness selected to accommodate various tolerances of the fuel cell elements and are either bonded to one of the other frames or to a bipolar or end plate. One of the non-compliant frames is wider than the other frames forming an overlap of the matrix over the wider seal frame, which cooperates with electrolyte permeating the matrix to form a wet seal within the fuel cell that prevents process gases from intermixing at the periphery of the fuel cell and a dry seal surrounding the cell to keep electrolyte from the periphery thereof. The frames may be made in one piece, in L-shaped portions or in strips and have an outer perimeter which registers with the outer perimeter of bipolar or end plates to form surfaces upon which flanges of pan shaped, gas manifolds can be sealed.

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConnell, Paul E.; Ross, Steven; Grey, Carissa Ann

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacificmore » Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.« less

IMPROVED TYPE OF FUEL ELEMENT

DOEpatents

Monson, H.O.

1961-01-24

A radiator-type fuel block assembly is described. It has a hexagonal body of neutron fissionable material having a plurality of longitudinal equal- spaced coolant channels therein aligned in rows parallel to each face of the hexagonal body. Each of these coolant channels is hexagonally shaped with the corners rounded and enlarged and the assembly has a maximum temperature isothermal line around each channel which is approximately straight and equidistant between adjacent channels.

Posttest examination of Sodium Loop Safety Facility experiments. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holland, J.W.

In-reactor, safety experiments performed in the Sodium Loop Safety Facility (SLSF) rely on comprehensive posttest examinations (PTE) to characterize the postirradiation condition of the cladding, fuel, and other test-subassembly components. PTE information and on-line instrumentation data, are analyzed to identify the sequence of events and the severity of the accident for each experiment. Following in-reactor experimentation, the SLSF loop and test assembly are transported to the Hot Fuel Examination Facility (HFEF) for initial disassembly. Goals of the HFEF-phase of the PTE are to retrieve the fuel bundle by dismantling the loop and withdrawing the test assembly, to assess the macro-conditionmore » of the fuel bundle by nondestructive examination techniques, and to prepare the fuel bundle for shipment to the Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory.« less

Horizontal modular dry irradiated fuel storage system

DOEpatents

Fischer, Larry E.; McInnes, Ian D.; Massey, John V.

1988-01-01

A horizontal, modular, dry, irradiated fuel storage system (10) includes a thin-walled canister (12) for containing irradiated fuel assemblies (20), which canister (12) can be positioned in a transfer cask (14) and transported in a horizontal manner from a fuel storage pool (18), to an intermediate-term storage facility. The storage system (10) includes a plurality of dry storage modules (26) which accept the canister (12) from the transfer cask (14) and provide for appropriate shielding about the canister (12). Each module (26) also provides for air cooling of the canister (12) to remove the decay heat of the irradiated fuel assemblies (20). The modules (26) can be interlocked so that each module (26) gains additional shielding from the next adjacent module (26). Hydraulic rams (30) are provided for inserting and removing the canisters (12) from the modules (26).

Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

NASA Astrophysics Data System (ADS)

Artnak, Edward Joseph, III

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

Southern phosphorus indices, water quality data, and modeling (APEX, APLE, and TBET) results: A comparison

USDA-ARS?s Scientific Manuscript database

Phosphorus (P) indices in the southern United States frequently produce different recommendations for similar conditions. After assembling data from benchmark southern sites (Arkansas, Georgia, Mississippi, North Carolina, Oklahoma, and Texas), land treatment information was used in the 12 southern...

FUEL ROD ASSEMBLY

DOEpatents

Hutter, E.

1959-09-01

A cluster of nuclear fuel rods aod a tubular casing through which a coolant flows in heat-change contact with the ruel rods are described. The casting is of trefoil section and carries the fuel rods, each of which has two fin engaging the serrated fins of the other two fuel rods, whereby the fuel rods are held in the casing and are interlocked against relative longitudinal movement.

MetaQUAST: evaluation of metagenome assemblies.

PubMed

Mikheenko, Alla; Saveliev, Vladislav; Gurevich, Alexey

2016-04-01

During the past years we have witnessed the rapid development of new metagenome assembly methods. Although there are many benchmark utilities designed for single-genome assemblies, there is no well-recognized evaluation and comparison tool for metagenomic-specific analogues. In this article, we present MetaQUAST, a modification of QUAST, the state-of-the-art tool for genome assembly evaluation based on alignment of contigs to a reference. MetaQUAST addresses such metagenome datasets features as (i) unknown species content by detecting and downloading reference sequences, (ii) huge diversity by giving comprehensive reports for multiple genomes and (iii) presence of highly relative species by detecting chimeric contigs. We demonstrate MetaQUAST performance by comparing several leading assemblers on one simulated and two real datasets. http://bioinf.spbau.ru/metaquast aleksey.gurevich@spbu.ru Supplementary data are available at Bioinformatics online. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com.

Analysis of the type II robotic mixed-model assembly line balancing problem

NASA Astrophysics Data System (ADS)

Çil, Zeynel Abidin; Mete, Süleyman; Ağpak, Kürşad

2017-06-01

In recent years, there has been an increasing trend towards using robots in production systems. Robots are used in different areas such as packaging, transportation, loading/unloading and especially assembly lines. One important step in taking advantage of robots on the assembly line is considering them while balancing the line. On the other hand, market conditions have increased the importance of mixed-model assembly lines. Therefore, in this article, the robotic mixed-model assembly line balancing problem is studied. The aim of this study is to develop a new efficient heuristic algorithm based on beam search in order to minimize the sum of cycle times over all models. In addition, mathematical models of the problem are presented for comparison. The proposed heuristic is tested on benchmark problems and compared with the optimal solutions. The results show that the algorithm is very competitive and is a promising tool for further research.

77 FR 2658 - Airworthiness Directives; Bombardier, Inc. Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-19

... high pressure fuel lines due to improper installation of an expandable pin on the lower cowl assembly... chafing of the high pressure fuel lines, which if not corrected, could cause fuel leakage in a fire zone... on a high pressure (HP) fuel line. The source of chafing was related to the improper installation of...

Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

DOEpatents

Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

2012-04-24

A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

Nuclear reactors built, being built, or planned, 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, B.

1992-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Improving Energy Efficiency for the Vehicle Assembly Industry: A Discrete Event Simulation Approach

NASA Astrophysics Data System (ADS)

Oumer, Abduaziz; Mekbib Atnaw, Samson; Kie Cheng, Jack; Singh, Lakveer

2016-11-01

This paper presented a Discrete Event Simulation (DES) model for investigating and improving energy efficiency in vehicle assembly line. The car manufacturing industry is one of the highest energy consuming industries. Using Rockwell Arena DES package; a detailed model was constructed for an actual vehicle assembly plant. The sources of energy considered in this research are electricity and fuel; which are the two main types of energy sources used in a typical vehicle assembly plant. The model depicts the performance measurement for process- specific energy measures of painting, welding, and assembling processes. Sound energy efficiency model within this industry has two-fold advantage: reducing CO2 emission and cost reduction associated with fuel and electricity consumption. The paper starts with an overview of challenges in energy consumption within the facilities of automotive assembly line and highlights the parameters for energy efficiency. The results of the simulation model indicated improvements for energy saving objectives and reduced costs.

Neutron Activation and Thermoluminescent Detector Responses to a Bare Pulse of the CEA Valduc SILENE Critical Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Thomas Martin; Celik, Cihangir; McMahan, Kimberly L.

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 11, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). Themore » goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.« less

Combustor assembly in a gas turbine engine

DOEpatents

Wiebe, David J; Fox, Timothy A

2013-02-19

A combustor assembly in a gas turbine engine. The combustor assembly includes a combustor device coupled to a main engine casing, a first fuel injection system, a transition duct, and an intermediate duct. The combustor device includes a flow sleeve for receiving pressurized air and a liner disposed radially inwardly from the flow sleeve. The first fuel injection system provides fuel that is ignited with the pressurized air creating first working gases. The intermediate duct is disposed between the liner and the transition duct and defines a path for the first working gases to flow from the liner to the transition duct. An intermediate duct inlet portion is associated with a liner outlet and allows movement between the intermediate duct and the liner. An intermediate duct outlet portion is associated with a transition duct inlet section and allows movement between the intermediate duct and the transition duct.

Isochoric Implosions for Fast Ignition

NASA Astrophysics Data System (ADS)

Clark, Daniel; Tabak, Max

2006-10-01

Various gain models have shown the potentially great advantages of Fast Ignition (FI) Inertial Confinement Fusion (ICF) over its conventional hotspot ignition counterpart. These gain models, however, all assume nearly uniform-density fuel assemblies. By contrast, typical ICF implosions yield hollowed fuel assemblies with a high-density shell of fuel surrounding a low-density, high-pressure hotspot. To realize fully the advantages of FI, then, an alternative implosion design must be found which yields nearly isochoric fuel assemblies without substantial hotspots. Here, it is shown that a self-similar spherical implosion of the type originally studied by Guderley [Luftfahrtforschung 19, 302 (1942)] may be employed to yield precisely such quasi-isochoric imploded states. The difficulty remains, however, of accessing these self-similarly imploding configurations from initial conditions representing an actual ICF target, namely a uniform, solid-density shell at rest. Furthermore, these specialized implosions must be realized for practicable drive parameters, i.e., accessible peak pressures, shell aspect ratios, etc. An implosion scheme is presented which meets all of these requirements, suggesting the possibility of genuinely isochoric implosions for FI.

Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

Single-wall carbon nanotube-based proton exchange membrane assembly for hydrogen fuel cells.

PubMed

Girishkumar, G; Rettker, Matthew; Underhile, Robert; Binz, David; Vinodgopal, K; McGinn, Paul; Kamat, Prashant

2005-08-30

A membrane electrode assembly (MEA) for hydrogen fuel cells has been fabricated using single-walled carbon nanotubes (SWCNTs) support and platinum catalyst. Films of SWCNTs and commercial platinum (Pt) black were sequentially cast on a carbon fiber electrode (CFE) using a simple electrophoretic deposition procedure. Scanning electron microscopy and Raman spectroscopy showed that the nanotubes and the platinum retained their nanostructure morphology on the carbon fiber surface. Electrochemical impedance spectroscopy (EIS) revealed that the carbon nanotube-based electrodes exhibited an order of magnitude lower charge-transfer reaction resistance (R(ct)) for the hydrogen evolution reaction (HER) than did the commercial carbon black (CB)-based electrodes. The proton exchange membrane (PEM) assembly fabricated using the CFE/SWCNT/Pt electrodes was evaluated using a fuel cell testing unit operating with H(2) and O(2) as input fuels at 25 and 60 degrees C. The maximum power density obtained using CFE/SWCNT/Pt electrodes as both the anode and the cathode was approximately 20% better than that using the CFE/CB/Pt electrodes.

Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Chang, G. S.; Lillo, M. A.

2009-08-01

The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y-Z mini-plate fuel model was developed. The Y-Z model divides each fuel plate into 30 equal intervals in both the Y and Z directions. The MCNP-calculated results and the detailed Y-Z fission power mapping were used to help design the AFIP fuel test assembly to demonstrate that the AFIP test assembly thermal-hydraulic limits will not exceed the ATR safety limits.

78 FR 27951 - Foreign-Trade Zone (FTZ) 75-Phoenix, Arizona; Notification of Proposed Production Activity...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-13

... systems; duct temperature limiters; air/oil heat exchangers; oil cooler fans; fuel filter assemblies... assemblies; filter extractors; de- coupler/disassembly wrenches; torque wrench adaptors; test benches; drills...; filter assemblies; oil filter install kits; cartridge screens; filter housings; trim balance weights...

A nine-scaffold genome assembly of the nine chromosome sugar beet

USDA-ARS?s Scientific Manuscript database

A sugar beet genome sequence is required to take full advantage of the increasingly powerful approaches directed a single nucleotide resolution across the whole genome. A high quality reference genome serves as a benchmark from which other genotypes might be compared and exploited for sugar beet imp...

Fuel nozzle assembly for use in turbine engines and methods of assembling same

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward

2015-02-03

A fuel nozzle for use with a turbine engine is described herein. The fuel nozzle includes a housing that is coupled to a combustor liner defining a combustion chamber. The housing includes an endwall that at least partially defines the combustion chamber. A plurality of mixing tubes extends through the housing for channeling fuel to the combustion chamber. Each mixing tube of the plurality of mixing tubes includes an inner surface that extends between an inlet portion and an outlet portion. The outlet portion is oriented adjacent the housing endwall. At least one of the plurality of mixing tubes includes a plurality of projections that extend outwardly from the outlet portion. Adjacent projections are spaced a circumferential distance apart such that a groove is defined between each pair of circumferentially-apart projections to facilitate enhanced mixing of fuel in the combustion chamber.

Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

NASA Astrophysics Data System (ADS)

Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

2016-08-01

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel assemblies. Based on analysis of modern engineering solutions on SNF regeneration, it has been concluded that new reprocessing technologies should meet the ecological safety requirements, provide a more extensive use of the resource base of nuclear engineering, allow the production of valuable and trace elements on an industrial scale, and decrease radioactive waste release.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

F. Delage; J. Carmack; C. B. Lee

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 oxidemore » 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.« less

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, Charles W.

1992-01-01

A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

Spacer grid assembly and locking mechanism

DOEpatents

Snyder, Jr., Harold J.; Veca, Anthony R.; Donck, Harry A.

1982-01-01

A spacer grid assembly is disclosed for retaining a plurality of fuel rods in substantially parallel spaced relation, the spacer grids being formed with rhombic openings defining contact means for engaging from one to four fuel rods arranged in each opening, the spacer grids being of symmetric configuration with their rhombic openings being asymmetrically offset to permit inversion and relative rotation of the similar spacer grids for improved support of the fuel rods. An improved locking mechanism includes tie bars having chordal surfaces to facilitate their installation in slotted circular openings of the spacer grids, the tie rods being rotatable into locking engagement with the slotted openings.

10 CFR 32.2 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Definitions. 32.2 Section 32.2 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL... disposal, or nuclear material contained in any fuel assembly, subassembly, fuel rod, or fuel pellet...

10 CFR 32.2 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Definitions. 32.2 Section 32.2 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL... disposal, or nuclear material contained in any fuel assembly, subassembly, fuel rod, or fuel pellet...

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Gurinsky, D.H.; Powell, R.W.; Fox, M.

1959-11-24

A nuclear fuel element comprising a plurality of nuclear fuel bearing strips is presented. The strips are folded along their longitudinal axes to an angle of about 60 deg and are secured at each end by ferrule to form an elongated assembly suitable for occupying a cylindrical coolant channel.

A historical examination of the nature of science and its consensus as presented in the Benchmarks for Science Literacy and National Science Education Standards

NASA Astrophysics Data System (ADS)

Felske, Daniel D.

Developing a scientific literate citizenry has fueled science education reforms for the past 40 years. A review of the literature reveals that definitions of scientific literacy during this period were greatly influenced by the goals, directions, and political agendas of the day. This approach has resulted in programs emphasizing certain aspects of scientific literacy while neglecting others. Additionally, consensus on what scientific literacy means or how to develop it has not been achieved. One aspect of scientific literacy that is agreed upon is the essential role that the nature of science (NOS) plays in its development. For this reason, an extensive review of the literature was conducted to develop a comprehensive background of this topic. The component structure of the NOS revealed in the literature was then synthesized into a NOS framework. The NOS framework served to guide the construction of a 21 item questionnaire taken from statements embedded in the consensual documents Benchmarks for Science Literacy (AAAS, 1993) and National Science Education Standards (NRC, 1996). A panel of five experts who have written extensively on the nature of science was then assembled and the degree of NOS consensus measured using a modified Delphi technique. The results of the survey indicated a high level of consensus (95%) at the ≥80% level. The panelists concurred positively on 19 of 21 NOS items, concurred negatively on one of 21 NOS items (item 10), and could not reach consensus on one of 21 NOS items (item 16). These findings, as well as, the NOS framework, are important first steps toward developing programs that foster the development of scientific literacy.

Upper Stage Flight Experiment 10K Engine Design and Test Results

NASA Technical Reports Server (NTRS)

Ross, R.; Morgan, D.; Crockett, D.; Martinez, L.; Anderson, W.; McNeal, C.

2000-01-01

A 10,000 lbf thrust chamber was developed for the Upper Stage Flight Experiment (USFE). This thrust chamber uses hydrogen peroxide/JP-8 oxidizer/fuel combination. The thrust chamber comprises an oxidizer dome and manifold, catalyst bed assembly, fuel injector, and chamber/nozzle assembly. Testing of the engine was done at NASA's Stennis Space Center (SSC) to verify its performance and life for future upper stage or Reusable Launch Vehicle applications. Various combinations of silver screen catalyst beds, fuel injectors, and combustion chambers were tested. Results of the tests showed high C* efficiencies (97% - 100%) and vacuum specific impulses of 275 - 298 seconds. With fuel film cooling, heating rates were low enough that the silica/quartz phenolic throat experienced minimal erosion. Mission derived requirements were met, along with a perfect safety record.

A review of the current state-of-the-art methodology for handling bias and uncertainty in performing criticality safety evaluations. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Disney, R.K.

1994-10-01

The methodology for handling bias and uncertainty when calculational methods are used in criticality safety evaluations (CSE`s) is a rapidly evolving technology. The changes in the methodology are driven by a number of factors. One factor responsible for changes in the methodology for handling bias and uncertainty in CSE`s within the overview of the US Department of Energy (DOE) is a shift in the overview function from a ``site`` perception to a more uniform or ``national`` perception. Other causes for change or improvement in the methodology for handling calculational bias and uncertainty are; (1) an increased demand for benchmark criticalsmore » data to expand the area (range) of applicability of existing data, (2) a demand for new data to supplement existing benchmark criticals data, (3) the increased reliance on (or need for) computational benchmarks which supplement (or replace) experimental measurements in critical assemblies, and (4) an increased demand for benchmark data applicable to the expanded range of conditions and configurations encountered in DOE site restoration and remediation.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hannan, N. A.; Matos, J. E.; Stillman, J. A.

At the request of the Czech Technical University (CTU) in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. Fore core configurations C1 and C2, criticality calculations were done for cases with all controlmore » rods at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were doe for the C1 core configuration. The reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations. Finally, the reactivity feedback coefficients, the prompt neutron lifetime, and the total effective delay neutron fraction were calculated for each of the three cores.« less

TREAT Transient Analysis Benchmarking for the HEU Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D. C.; Connaway, H. M.; Wright, A. E.

2014-05-01

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used tomore » determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.« less

Advanced propulsion engine assessment based on a cermet reactor

NASA Technical Reports Server (NTRS)

Parsley, Randy C.

1993-01-01

A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruggles, A.E.

1990-10-12

The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results aremore » related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.« less

PEM/SPE fuel cell

DOEpatents

Grot, Stephen Andreas

1998-01-01

A PEM/SPE fuel cell including a membrane-electrode assembly (MEA) having a plurality of oriented filament embedded the face thereof for supporting the MEA and conducting current therefrom to contiguous electrode plates.

Method of improving fuel cell performance by removing at least one metal oxide contaminant from a fuel cell electrode

DOEpatents

Kim, Yu Seung [Los Alamos, NM; Choi, Jong-Ho [Los Alamos, NM; Zelenay, Piotr [Los Alamos, NM

2009-08-18

A method of removing contaminants from a fuel cell catalyst electrode. The method includes providing a getter electrode and a fuel cell catalyst electrode having at least one contaminant to a bath and applying a voltage sufficient to drive the contaminant from the fuel cell catalyst electrode to the getter electrode. Methods of removing contaminants from a membrane electrode assembly of a fuel cell and of improving performance of a fuel cell are also provided.

Benchmarking study of the MCNP code against cold critical experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, S.

1991-01-01

The purpose of this study was to benchmark the widely used Monte Carlo code MCNP against a set of cold critical experiments with a view to using the code as a means of independently verifying the performance of faster but less accurate Monte Carlo and deterministic codes. The experiments simulated consisted of both fast and thermal criticals as well as fuel in a variety of chemical forms. A standard set of benchmark cold critical experiments was modeled. These included the two fast experiments, GODIVA and JEZEBEL, the TRX metallic uranium thermal experiments, the Babcock and Wilcox oxide and mixed oxidemore » experiments, and the Oak Ridge National Laboratory (ORNL) and Pacific Northwest Laboratory (PNL) nitrate solution experiments. The principal case studied was a small critical experiment that was performed with boiling water reactor bundles.« less

Resource requirements of inclusive urban development in India: insights from ten cities

NASA Astrophysics Data System (ADS)

Singh Nagpure, Ajay; Reiner, Mark; Ramaswami, Anu

2018-02-01

This paper develops a methodology to assess the resource requirements of inclusive urban development in India and compares those requirements to current community-wide material and energy flows. Methods include: (a) identifying minimum service level benchmarks for the provision of infrastructure services including housing, electricity and clean cooking fuels; (b) assessing the percentage of homes that lack access to infrastructure or that consume infrastructure services below the identified benchmarks; (c) quantifying the material requirements to provide basic infrastructure services using India-specific design data; and (d) computing material and energy requirements for inclusive development and comparing it with current community-wide material and energy flows. Applying the method to ten Indian cities, we find that: 1%-6% of households do not have electricity, 14%-71% use electricity below the benchmark of 25 kWh capita-month-1 4%-16% lack structurally sound housing; 50%-75% live in floor area less than the benchmark of 8.75 m2 floor area/capita; 10%-65% lack clean cooking fuel; and 6%-60% lack connection to a sewerage system. Across the ten cities examined, to provide basic electricity (25 kWh capita-month-1) to all will require an addition of only 1%-10% in current community-wide electricity use. To provide basic clean LPG fuel (1.2 kg capita-month-1) to all requires an increase of 5%-40% in current community-wide LPG use. Providing permanent shelter (implemented over a ten year period) to populations living in non-permanent housing in Delhi and Chandigarh would require a 6%-14% increase over current annual community-wide cement use. Conversely, to provide permanent housing to all people living in structurally unsound housing and those living in overcrowded housing (<5 m cap-2) would require 32%-115% of current community-wide cement flows. Except for the last scenario, these results suggest that social policies that seek to provide basic infrastructure provisioning for all residents would not dramatically increasing current community-wide resource flows.

Self-actuated nuclear reactor shutdown system using induction pump to facilitate sensing of core coolant temperature

DOEpatents

Sievers, Robert K.; Cooper, Martin H.; Tupper, Robert B.

1987-01-01

A self-actuated shutdown system incorporated into a reactivity control assembly in a nuclear reactor includes pumping means for creating an auxiliary downward flow of a portion of the heated coolant exiting from the fuel assemblies disposed adjacent to the control assembly. The shutdown system includes a hollow tubular member which extends through the outlet of the control assembly top nozzle so as to define an outer annular flow channel through the top nozzle outlet separate from an inner flow channel for primary coolant flow through the control assembly. Also, a latching mechanism is disposed in an inner duct of the control assembly and is operable for holding absorber bundles in a raised position in the control assembly and for releasing them to drop them into the core of the reactor for shutdown purposes. The latching mechanism has an inner flow passage extending between and in flow communication with the absorber bundles and the inner flow channel of the top nozzle for accommodating primary coolant flow upwardly through the control assembly. Also, an outer flow passage separate from the inner flow passage extends through the latching mechanism between and in flow communication with the inner duct and the outer flow channel of the top nozzle for accommodating inflow of a portion of the heated coolant from the adjacent fuel assemblies. The latching mechanism contains a magnetic material sensitive to temperature and operable to cause mating or latching together of the components of the latching mechanism when the temperature sensed is below a known temperature and unmating or unlatching thereof when the temperature sensed is above a given temperature. The temperature sensitive magnetic material is positioned in communication with the heated coolant flow through the outer flow passage for directly sensing the temperature thereof. Finally, the pumping means includes a jet induction pump nozzle and diffuser disposed adjacent the bottom nozzle of the control assembly and in flow communication with the inlet thereof. The pump nozzle is operable to create an upward driving flow of primary coolant through the pump diffuser and then to the absorber bundles. The upward driving flow of primary coolant, in turn, creates a suction head within the outer flow channel of the top nozzle and thereby an auxiliary downward flow of the heated coolant portion exiting from the upper end of the adjacent fuel assemblies through the outer flow channel to the pump nozzle via the outer flow passage of the latching mechanism and an annular space between the outer and inner spaced ducts of the control assembly housing. The temperature of the heated coolant exiting from the adjacent fuel assemblies can thereby be sensed directly by the temperature sensitive magnetic material in the latching mechanism.

78 FR 5773 - Notification of Proposed Production Activity, Generac Power Systems, Inc., Subzone 41J...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-28

... assemblies; oil/fuel filters; air/oil separation equipment; air filters/elements; catalytic converters... assemblies; AC line filters; dielectric items of paper/plastic; capacitors; circuit breakers; switching...