DoD Fuel Facilities Criteria

DTIC Science & Technology

2015-04-27

Pantograph Feb-2010 UFGS 33 58 00 Leak Detection for Fueling Systems Apr-2008 UFGS 33 52 43.13 Aviation Fuel Piping Feb-2010 UFGS 33 59 00 Tightness of... Pipeline Pressure Testing Guidelines  Specifications  Questions 2 7/12/2017 3 7/12/2017 DoD Fuels Facilities Documents  Unified...UFGS)  Most in the 33 nn nn series  Associated with Standard Designs  Available on WBDG site  Coating Systems 4 7/12/2017 Pipeline

Multi-dimensional transport modelling of corrosive agents through a bentonite buffer in a Canadian deep geological repository.

PubMed

Briggs, Scott; McKelvie, Jennifer; Sleep, Brent; Krol, Magdalena

2017-12-01

The use of a deep geological repository (DGR) for the long-term disposal of used nuclear fuel is an approach currently being investigated by several agencies worldwide, including Canada's Nuclear Waste Management Organization (NWMO). Within the DGR, used nuclear fuel will be placed in copper-coated steel containers and surrounded by a bentonite clay buffer. While copper is generally thermodynamically stable, corrosion can occur due to the presence of sulphide under anaerobic conditions. As such, understanding transport of sulphide through the engineered barrier system to the used fuel container is an important consideration in DGR design. In this study, a three-dimensional (3D) model of sulphide transport in a DGR was developed. The numerical model is implemented using COMSOL Multiphysics, a commercial finite element software package. Previous sulphide transport models of the NWMO repository used a simplified one-dimensional system. This work illustrates the importance of 3D modelling to capture non-uniform effects, as results showed locations of maximum sulphide flux are 1.7 times higher than the average flux to the used fuel container. Copyright © 2017. Published by Elsevier B.V.

Research on Fire-Resistant Diesel Fuel.

DTIC Science & Technology

1981-12-01

1[ >1 1? 1F RESEARCH ON FIRE-RESISTANT DIESEL FUEL INTERIM REPORT AFLRL No. 145 By W.D. Weatherford, Jr. G.E. Fodor M.D. Kanakia D.W. Naegeli B.R...GRANT NUMBER(S) W.D. Weatherford, Jr., G.E. Fodor, DAAK7O-80-C-OOO1 M.D. Kanakia, D.W. Naegeli , B.R. Wright (AFLRL) DAAK7O-82-C-OOO1 andF.W...Filters in Fuel Systems of U.S. Army Vehicles I:I , i , , , =- 111 REFERENCES 1. Weatherford, W.D., Jr. Fodor, G.E., Naegeli , D.W., Owens, E.C., Wright

Propulsion and Power Rapid Response Research and Development (R&D) Support. Delivery Order 0011: Advanced Propulsion Fuels R&D, Subtask: Evaluation of 50/50 Hydroprocessed Renewable Jet Fuel and JP8 in the Ford 6.7L High-Pressure Common Rail Design Engine

DTIC Science & Technology

2012-12-01

treated with 9ppm of a QPL-25017 additive. Testing was completed using a Ford 6.7L V8 turbocharged diesel engine. Testing was completed following a...Installation ................................................................................. 3 2. Ford 6.7L Fuel Injection Pump, Rail, & Injector ...5 4. Fuel Injector Component Break-Out

Production test IP-376-D, Supplement B Irradiation of MGCR-HDR-3 Test Element

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

Baars, R.E.

The objective of this supplement to PT-IP-376-D, Irradiation of MGCR-HDR-3 Test Element is to authorize 1000 hours of operation at a maximum test specimen surface temperature of 1700 F. The original production test authorized a test duration of four months at a maximum specimen surface temperature of 1500 F; supplement A authorized extension of the test duration to ten months. The desired increase in surface temperature is requested to demonstrate the general feasibility of operation of the fuel element at 1700 F, and to obtain specific information on the performance of Hastelloy-X cladding and fuel bodies. The increased temperature hasmore » been approved by the Atomic Energy Commission.« less

Fuel clad chemical interactions in fast reactor MOX fuels

NASA Astrophysics Data System (ADS)

Viswanathan, R.

2014-01-01

Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel-Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ṡ [B/(at.% fission)] ṡ (T/K-705) ṡ [(O/M)i-1.935]} + 20.5) for (O/M)i ⩽ 1.98. A new model is proposed for (O/M)i ⩾ 1.98: d/μm = [B/(at.% fission)] ṡ (T/K-800)0.5 ṡ [(O/M)i-1.94] ṡ [P/(W cm-1)]0.5. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M)i is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

Fuel Composition Analysis of Endothermically Heated JP-8 Fuel for Use in a Pulse Detonation Engine

DTIC Science & Technology

2008-06-01

detonation engine (PDE) was extracted via zeolite catalyst coated concentric tube-counter flow heat exchangers to produce supercritical pyrolytic conditions...gas chromatography flame ionization and thermal conductivity detectors ............................................. 68 Table B.1. Elemental bias... chromatography ...................... 98 Table D.1b. Products found in the liquid sample by gas chromatography (continued) ... 99 Table D.1c

Commercial Approval Plan for Synthetic Jet Fuel from Hydrotreated Fats and Oils

DTIC Science & Technology

2009-02-18

driven by their experience, some of it very negative, with the other more well known organic oil derived fuel, BioDiesel. BioDiesel is methyl ester of...the fatty acid ( FAME ) that comes from the triglycerides that compose the organic oil. The HRJ SPKs are deoxygenated materials that are processed in...SwRI Cu PE506 * Semi-Quant Survey ICP/MS * Organic Elements C:H D5291 * N D4629 * S D5453 * Acid Number D3242 * Carbonyls, alcohols, esters , phenols

Toxicity of irradiated advanced heavy water reactor fuels.

PubMed

Priest, N D; Richardson, R B; Edwards, G W R

2013-02-01

The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

A Variational Nodal Approach to 2D/1D Pin Resolved Neutron Transport for Pressurized Water Reactors

DOE PAGES

Zhang, Tengfei; Lewis, E. E.; Smith, M. A.; ...

2017-04-18

A two-dimensional/one-dimensional (2D/1D) variational nodal approach is presented for pressurized water reactor core calculations without fuel-moderator homogenization. A 2D/1D approximation to the within-group neutron transport equation is derived and converted to an even-parity form. The corresponding nodal functional is presented and discretized to obtain response matrix equations. Within the nodes, finite elements in the x-y plane and orthogonal functions in z are used to approximate the spatial flux distribution. On the radial interfaces, orthogonal polynomials are employed; on the axial interfaces, piecewise constants corresponding to the finite elements eliminate the interface homogenization that has been a challenge for method ofmore » characteristics (MOC)-based 2D/1D approximations. The angular discretization utilizes an even-parity integral method within the nodes, and low-order spherical harmonics (P N) on the axial interfaces. The x-y surfaces are treated with high-order P N combined with quasi-reflected interface conditions. Furthermore, the method is applied to the C5G7 benchmark problems and compared to Monte Carlo reference calculations.« less

A Variational Nodal Approach to 2D/1D Pin Resolved Neutron Transport for Pressurized Water Reactors

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

Zhang, Tengfei; Lewis, E. E.; Smith, M. A.

A two-dimensional/one-dimensional (2D/1D) variational nodal approach is presented for pressurized water reactor core calculations without fuel-moderator homogenization. A 2D/1D approximation to the within-group neutron transport equation is derived and converted to an even-parity form. The corresponding nodal functional is presented and discretized to obtain response matrix equations. Within the nodes, finite elements in the x-y plane and orthogonal functions in z are used to approximate the spatial flux distribution. On the radial interfaces, orthogonal polynomials are employed; on the axial interfaces, piecewise constants corresponding to the finite elements eliminate the interface homogenization that has been a challenge for method ofmore » characteristics (MOC)-based 2D/1D approximations. The angular discretization utilizes an even-parity integral method within the nodes, and low-order spherical harmonics (P N) on the axial interfaces. The x-y surfaces are treated with high-order P N combined with quasi-reflected interface conditions. Furthermore, the method is applied to the C5G7 benchmark problems and compared to Monte Carlo reference calculations.« less

77 FR 34367 - Proposed Subsequent Arrangement

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-11

... reactors, and a research reactor, at the Post Irradiation Examination Facility (PIEF), the Irradiated.../2011, ``Post-Irradiation Examination and R&D Programs Using Irradiated Fuels at KAERI,'' dated June... fuel elements for post-irradiation examination and for research, development and manufacture of DUPIC...

Sodium aluminum-iron phosphate glass-ceramics for immobilization of lanthanide oxide wastes from pyrochemical reprocessing of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Stefanovsky, S. V.; Stefanovsky, O. I.; Kadyko, M. I.; Nikonov, B. S.

2018-03-01

Sodium aluminum (iron) phosphate glass ceramics containing of up to 20 wt.% rare earth (RE) oxides simulating pyroprocessing waste were produced by melting at 1250 °C followed by either quenching or slow cooling to room temperature. The iron-free glass-ceramics were composed of major glass and minor phosphotridymite and monazite. The iron-bearing glass-ceramics were composed of major glass and minor monazite and Na-Al-Fe orthophosphate at low waste loadings (5-10 wt.%) and major orthophosphate and minor monazite as well as interstitial glass at high waste loadings (15-20 wt.%). Slowly cooled samples contained higher amount of crystalline phases than quenched ones. Monazite is major phase for REs. Leach rates from the materials of major elements (Na, Al, Fe, P) are 10-5-10-7 g cm-2 d-1, RE elements - lower than 10-5 g cm-2 d-1.

Issues and Potential Program on Denatured Fuel Utilization.

DTIC Science & Technology

1978-12-01

HTGR fuel develop - ment program ; 4. coated particles of (U,Th)02 have been extensively tested as potential HTGR fuels . A detailed summary of the...current scrap and waste treatment requirements. dBase case for all HTGR (Prismatic Fuel Element) cases based on data in "Summary Program Plan...Alternate Program for HTGR Fuel Recycle," April 11, 1975, Draft. 19 a --- AC8NCi09 The principal factors that result in a nominally-higher cost for

Radiation effect of neutrons produced by D-D side reactions on a D-3He fusion reactor

NASA Astrophysics Data System (ADS)

Bahmani, J.

2017-04-01

One of the most important characteristics in D-3He fusion reactors is neutron production via D-D side reactions. The neutrons can activate structural material, degrading them and ultimately converting them into high-level radioactive waste, while it is really costly and difficult to remove them. The neutrons from a fusion reactor could also be used to make weapons-grade nuclear material, rendering such types of fusion reactors a serious proliferation hazard. A related problem is the presence of radioactive elements such as tritium in D-3He plasma, either as fuel for or as products of the nuclear reactions; substantial quantities of radioactive elements would not only pose a general health risk, but tritium in particular would also be another proliferation hazard. The problems of neutron radiation and radioactive element production are especially interconnected because both would result from the D-D side reaction. Therefore, the presentation approach for reducing neutrons via D-D nuclear side reactions in a D-3He fusion reactor is very important. For doing this research, energy losses and neutron power fraction in D-3He fusion reactors are investigated. Calculations show neutrons produced by the D-D nuclear side reaction could be reduced by changing to a more 3He-rich fuel mixture, but then the bremsstrahlung power loss fraction would increase in the D-3He fusion reactor.

Multidimensional Fuel Performance Code: BISON

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

BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phasemore » field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.« less

Multidimensional Multiphysics Simulation of TRISO Particle Fuel

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

J. D. Hales; R. L. Williamson; S. R. Novascone

2013-11-01

Multidimensional multiphysics analysis of TRISO-coated particle fuel using the BISON finite-element based nuclear fuels code is described. The governing equations and material models applicable to particle fuel and implemented in BISON are outlined. Code verification based on a recent IAEA benchmarking exercise is described, and excellant comparisons are reported. Multiple TRISO-coated particles of increasing geometric complexity are considered. It is shown that the code's ability to perform large-scale parallel computations permits application to complex 3D phenomena while very efficient solutions for either 1D spherically symmetric or 2D axisymmetric geometries are straightforward. Additionally, the flexibility to easily include new physical andmore » material models and uncomplicated ability to couple to lower length scale simulations makes BISON a powerful tool for simulation of coated-particle fuel. Future code development activities and potential applications are identified.« less

NORTHROP REACTOR. REVISION NO. 1 TO THE FINAL SAFEGUARDS REPORT

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

Duncan, J.M.; Shimizu, B.; Romine, R.A.

1962-10-01

Additions and changes related to the original application concerning construction and operation of the Northrop Reactor are given. Answers to 13 questions relative to the Final Safeguards Report are included. Answers are also included concerning 12 questions regarding receipt, possession, and storage of fuel elements. Other information is included concerning personnel changes and changes from Al-clad fuel elements to improved stainless steelclad hydride elements. It is concluded that the improved Northrop Reactor does not present any undue hazard to the health and safety of the operating personnel or the public. (J.R.D.)

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

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

Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings onmore » avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.« less

Acute Dermal Irritation Study of Six Jet Fuels in New Zealand White Rabbits: Comparison of Four Bio-Based Jet Fuels with Two Petroleum JP-8 Fuels

DTIC Science & Technology

2014-02-01

NA 5c. PROGRAM ELEMENT NUMBER 62202F 6. AUTHOR(S) Sterner, Teresa R.1; Hurley, Jonathon M.2; Edwards, James T.3; Shafer, Linda M.4; Mattie , David R... Mattie , D.R. 2014. Acute Dermal Irritation Study of Ten Jet Fuels in New Zealand White Rabbits: Comparison of Synthetic and Bio -Based Jet Fuels with...AFRL-RH-WP-TR-2014-0046 ACUTE DERMAL IRRITATION STUDY OF SIX JET FUELS IN NEW ZEALAND WHITE RABBITS: COMPARISON OF FOUR BIO -BASED JET FUELS

An assessment of the benefits of the use of NASA developed fuel conservative technology in the US commercial aircraft fleet

NASA Technical Reports Server (NTRS)

1975-01-01

Cost and benefits of a fuel conservative aircraft technology program proposed by NASA are estimated. NASA defined six separate technology elements for the proposed program: (a) engine component improvement (b) composite structures (c) turboprops (d) laminar flow control (e) fuel conservative engine and (f) fuel conservative transport. There were two levels postulated: The baseline program was estimated to cost $490 million over 10 years with peak funding in 1980. The level two program was estimated to cost an additional $180 million also over 10 years. Discussions with NASA and with representatives of the major commercial airframe manufacturers were held to estimate the combinations of the technology elements most likely to be implemented, the potential fuel savings from each combination, and reasonable dates for incorporation of these new aircraft into the fleet.

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

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

Melissa C. Teague; Brian P. Gorman; Steven L. Hayes

2013-10-01

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

First Annual Fuel Cell End Users Forum

DTIC Science & Technology

2011-05-03

UNCLASSIFIED: Dist A. Approved for public release 1 Conference Report First Annual Fuel Cell End Users Forum ...Report First Annual Fuel Cell End Users Forum 5a. CONTRACT NUMBER w56hzv-09-D-0154 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S...End Users Forum was established to provide a user focused venue that enables a peer-to-peer support network to exchange information and gain knowledge

DESIGN CRITERIA FOR FUEL DISSOLUTION SYSTEMS AND ASSOCIATED SERVICE FACILITIES. PLANT MODIFICATIONS FOR REPROCESSING NON-PRODUCTION REACTOR FUELS. PROJECT CGC-830

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

Bierman, S.R.; Graf, W.A.; Kass, M.

1960-07-29

Design panameters are presented for phases of the facility to reprocess low-enrichment fuels from nonproduction reactors. Included are plant flowsheets and equipment layouts for fuel element dissolution, centrifugation, solution adjustment, and waste handling. Also included are the basic design criteria for the supporting facilities which service these phases and all other facilites located in the vicinity of the selected building (Bldg. 221-U). (J.R.D.)

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

Burke, Timothy P.; Martz, Roger L.; Kiedrowski, Brian C.

New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructivemore » Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.« less

Solid oxide fuel cells, and air electrode and electrical interconnection materials therefor

DOEpatents

Bates, J. Lambert

1992-01-01

In one aspect of the invention, an air electrode material for a solid oxide fuel cell comprises Y.sub.1-a Q.sub.a MnO.sub.3, where "Q" is selected from the group consisting of Ca and Sr or mixtures thereof and "a" is from 0.1 to 0.8. Preferably, "a" is from 0.4 to 0.7. In another aspect of the invention, an electrical interconnection material for a solid oxide fuel cell comprises Y.sub.1-b Ca.sub.b Cr.sub.1-c Al.sub.c O.sub.3, where "b" is from 0.1 to 0.6 and "c" is from 0 to 9.3. Preferably, "b" is from 0.3 to 0.5 and "c" is from 0.05 to 0.1. A composite solid oxide electrochemical fuel cell incorporating these materials comprises: a solid oxide air electrode and an adjacent solid oxide electrical interconnection which commonly include the cation Y, the air electrode comprising Y.sub.1-a Q.sub.a MnO.sub.3, where "Q" is selected from the group consisting of Ca and Sr or mixtures thereof and "a" is from 0.1 to 0.8, the electrical interconnection comprising Y.sub.1-b Ca.sub.b Cr.sub.1-c Al.sub.c O.sub.3, where "b" is from 0.1 to 0.6 and "c" is from 0.0 to 0.3; a yttrium stabilized solid electrolyte comprising (1-d)ZrO.sub.2 -(d)Y.sub.2 O.sub.3 where "d" is from 0.06 to 0.5; and a solid fuel electrode comprising X-ZrO.sub.2, where "X" is an elemental metal.

Solid oxide fuel cells, and air electrode and electrical interconnection materials therefor

DOEpatents

Bates, J.L.

1992-09-01

In one aspect of the invention, an air electrode material for a solid oxide fuel cell comprises Y[sub 1[minus]a]Q[sub a]MnO[sub 3], where Q is selected from the group consisting of Ca and Sr or mixtures thereof and a' is from 0.1 to 0.8. Preferably, a' is from 0.4 to 0.7. In another aspect of the invention, an electrical interconnection material for a solid oxide fuel cell comprises Y[sub 1[minus]b]Ca[sub b]Cr[sub 1[minus]c]Al[sub c]O[sub 3], where b' is from 0.1 to 0.6 and c' is from 0 to 9.3. Preferably, b' is from 0.3 to 0.5 and c' is from 0.05 to 0.1. A composite solid oxide electrochemical fuel cell incorporating these materials comprises: a solid oxide air electrode and an adjacent solid oxide electrical interconnection which commonly include the cation Y, the air electrode comprising Y[sub 1[minus]a]Q[sub a]MnO[sub 3], where Q is selected from the group consisting of Ca and Sr or mixtures thereof and a' is from 0.1 to 0.8, the electrical interconnection comprising Y[sub 1[minus]b]Ca[sub b]Cr[sub 1[minus]c]Al[sub c]O[sub 3], where b' is from 0.1 to 0.6 and c' is from 0.0 to 0.3; a yttrium stabilized solid electrolyte comprising (1[minus]d)ZrO[sub 2]-(d)Y[sub 2]O[sub 3] where d' is from 0.06 to 0.5; and a solid fuel electrode comprising X-ZrO[sub 2], where X' is an elemental metal. 5 figs.

Overview of past and current activities on fuels for fast reactors at the Institute for Transuranium Elements

NASA Astrophysics Data System (ADS)

Fernandez, A.; McGinley, J.; Somers, J.; Walter, M.

2009-07-01

Nuclear energy has the potential to provide a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gas emissions. The renewal of interest in fast neutron spectra reactors to meet more ambitious sustainable development criteria (i.e., resource maximisation and waste minimisation), opens a favourable framework for R&D activities in this area. The Institute for Transuranium Elements has extensive experience in the fabrication, characterization and irradiation testing (Phénix, Dounreay, Rapsodie) of fast reactor fuels, in oxide, nitride and carbide forms. An overview of these past and current activities on fast reactor fuels is presented.

Fuel Property Effects on Diesel Engine and Gas Turbine Combustor Performance.

DTIC Science & Technology

1981-11-01

REPORT AFLRL No. 149 By A.F. Montemayor D.W. Naegeli -* L.G. Dodge E.C. Owens J.N. Bowden U.S. Army Fuels and Lubricants Research Laboratory Southwest...6800-120/ 1 7 AUTHOR18) S. CONTRACT ORt GRANT NUMBER(a) A.F. Vmotemeyor E. C. Owens DMA7-8O-C-0001 D. W. Naegeli J.N. Bowden DAAK70-82-C-OOO1 L.G...Acquisition Magazine, 18-20, September-October 1980. 5. Noses, C.A. and Naegeli , D.W., "Fuel Property Effects on Combustor Performance," ASME 79-GT-178

Chemical characterization of biomass fuel smoke particles of rural kitchens of South Asia

NASA Astrophysics Data System (ADS)

Deka, Pratibha; Hoque, Raza Rafiqul

2015-05-01

Biomass fuel smoke particles (BFSPs) of rural kitchens collected during dry and wet seasons were characterized for elements, anions and carbon. The BFSPs of kitchens using varied biomass fuel types viz. cow dung stick, mixed biomass, cow-dung stick-mixed biomass and sugarcane bagasse were chosen for the study. The BFSPs from cow dung fuel stick showed higher levels of elements, anions and particulate carbon than other BFSPs. Calcium, K, Fe and Mg were the major elements found in all BFSPs, which did not vary much between the seasons. Sulphate was found to be the dominant anion present in all BFSPs followed by Clˉ and PO43-. Seasonal variation was pronounced in the case of abundance of anions and particulate carbon. The ratio OC/EC, often used as source signature of biomass burning, was found to be within 1.89-7.41 and 1.72-6.19 during dry and wet seasons respectively.

Impact of Friction Reduction Technologies on Fuel Economy for Ground Vehicles

DTIC Science & Technology

2009-08-13

UNCLAS: Dist A. Approved for public release IMPACT OF FRICTION REDUCTION TECHNOLOGIES ON FUEL ECONOMY FOR GROUND VEHICLES G. R. Fenske , R. A. Erck...PROGRAM ELEMENT NUMBER 6. AUTHOR(S) G.R. Fenske ; R.A. Erck; O.O. Ajayi; A. Masoner’ A.S. Confort 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT

40 CFR 86.113-04 - Fuel specifications.

Code of Federal Regulations, 2012 CFR

2012-07-01

..., Research, Min. D 2699 93 Sensitivity, Min. 7.5 Lead (organic), max. g/U.S. gal. (g/liter) D 3237 0.050 (0... Retail octane number above the lowest octane rating that meets the fuel grade the manufacturer will... Retail octane number rather than a fuel grade, then the octane rating of the service accumulation...

Verification and Validation of the BISON Fuel Performance Code for PCMI Applications

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

Gamble, Kyle Allan Lawrence; Novascone, Stephen Rhead; Gardner, Russell James

2016-06-01

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described. Validation for application to light water reactor (LWR) PCMI problems is assessed by comparing predicted and measured rod diameter following base irradiation andmore » power ramps. Results indicate a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

Evaluation of Triethylene Glycol Monomethyl Ether (TRIEGME) as an Alternative Fuel System Icing Inhibitor for JP-8 Fuel

DTIC Science & Technology

2010-08-01

ALTERNATIVE FUEL SYSTEM ICING INHIBITOR FOR JP-8 FUEL 5a. CONTRACT NUMBER F33615-03-2-2347 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 62203F...Government. The authors would like to acknowledge funding support from the DoD Reduction of Total Ownership Cost program through Ed Wells of ASC...following individuals contributed substantially to the success of this program : Rex Cash of the 540 ACSS/GFLBB, Travis Whitmer of Boeing IDS, Tedd Biddle of

Artificial Leaf Based on Artificial Photosynthesis for Solar Fuel Production

DTIC Science & Technology

2017-06-30

AFRL-AFOSR-JP-TR-2017-0054 Artificial Leaf Based on Artificial Photosynthesis for Solar Fuel Production Mamoru Nango NAGOYA INSTITUTE OF TECHNOLOGY...for Solar Fuel Production 5a.  CONTRACT NUMBER 5b.  GRANT NUMBER FA2386-14-1-4015 5c.  PROGRAM ELEMENT NUMBER 61102F 6. AUTHOR(S) Mamoru Nango 5d...density immobilization of the photoreactants in the nanocavity inside PGP. The maximum production efficiency of formic acid inside the nanocavity was

Parametric Study of Advanced Mixing of Fuel/Oxidant System in High Speed Gaseous Flows and Experimental Validation Planning

DTIC Science & Technology

2001-08-30

Body with Thermo-Chemical destribution of Heat-Protected System . In: Physical and Gasdynamic Phenomena in Supersonic Flows Over Bodies. Edit. By...Final Report on ISTC Contract # 1809p Parametric Study of Advanced Mixing of Fuel/Oxidant System in High Speed Gaseous Flows and Experimental...of Advanced Mixing of Fuel/Oxidant System in High Speed Gaseous Flows and Experimental Validation Planning 5c. PROGRAM ELEMENT NUMBER 5d. PROJECT

The characteristics of bioethanol fuel made of vegetable raw materials

NASA Astrophysics Data System (ADS)

Muhaji; Sutjahjo, D. H.

2018-01-01

The aim of this research is to identify the most potential vegetable raw as the material to make a bioethanol fuel as the alternative energy for gasoline. This study used experimental method. The high-level bioethanol was obtained through the process of saccharification, fermentation and stratified distillation. ASTM standards were used as the method of testing the chemical element (D 5501, D 1744, D 1688, D 512, D 2622, D 381), and physical test (D 1613, D 240, D 1298-99, D 445, and D 93). The result of the analysis showed that from the seven bioethanols being studied there is one bioethanol from Saccharum of icinarum linn that has physical and chemical properties close to the standard of bioethanol. Meanwhile, the others only meet some of the physical and chemical properties of the standard bioethanol.

BISON Theory Manual The Equations behind Nuclear Fuel Analysis

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

Hales, J. D.; Williamson, R. L.; Novascone, S. R.

2016-09-01

BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact,more » and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.« less

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

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

Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

2015-09-01

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirementsmore » for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.« less

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)

Monitoring arrangement for vented nuclear fuel elements

DOEpatents

Campana, Robert J.

1981-01-01

In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

U.S.-Australia Civilian Nuclear Cooperation: Issues for Congress

DTIC Science & Technology

2010-09-30

7 Uranium Mining and Milling ................................................................................................8...cycle begins with mining uranium ore and upgrading it to yellowcake. Because naturally occurring uranium lacks sufficient fissile 235U to make fuel for...enrichment, and finally fabrication into fuel elements. Australia exports its uranium after the mining and milling stage. Commercial enrichment services

Dissolution of Used Nuclear Fuel Using a TBP/N-Paraffin Solvent

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

Rudisill, T. S.; Shehee, T. C.; Jones, D. H.

2017-10-02

The dissolution of unirradiated used nuclear fuel (UNF) pellets pretreated for tritium removal was demonstrated using a tributly phosphate (TBP) solvent. Dissolution of pretreated fuel in TBP could potentially combine dissolution with two cycle of solvent extraction required for separating the actinides and lanthanides from other fission products. Dissolutions were performed using UNF surrogates prepared from both uranyl nitrate and uranium trioxide produced from the pretreatment process by adding selected actinide and stable fission product elements. In laboratory-scale experiments, the U dissolution efficiency ranged from 80-99+% for both the nitrate and oxide surrogate fuels. On average, 80% of the Pumore » and 50% of the Np and Am in the nitrate surrogate dissolved; however, little of the transuranic elements dissolved in the oxide form. The majority of the 3+ lanthanide elements dissolved. Only small amounts of Sr (0-1.6%) and Mo (0.1-1.7%) and essentially no Cs, Ru, Zr, or Pd dissolved.« less

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

DOEpatents

Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

1999-03-23

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

Correlation of RP-1 Fuel Properties with Chemical Composition using Two-Dimensional Gas Chromatography with Time-of-Flight Mass Spectrometry followed by Partial Least Squares Regression Analysis

DTIC Science & Technology

2013-09-01

Striebich, S . P. Bagley , M. J. Wornant, Ind. Eng. Chem. Res. 50 (2011) 10434. [5] M. C. Billingsley, J. T. Edwards, L. M. Shafer, T. J. Bruno, AIAA 2010...CONTRACT NUMBER In-House 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR( S ) Kehimkar, B., J. Hoggard, L. Marney, M. Billingsley, C...Fraga, T. Bruno, and R. Synovec 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER Q0A4 7. PERFORMING ORGANIZATION NAME( S ) AND ADDRESS(ES

The Liquid Annular Reactor System (LARS) propulsion

NASA Technical Reports Server (NTRS)

Powell, James; Ludewig, Hans; Horn, Frederick; Lenard, Roger

1990-01-01

A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5).

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

NASA Astrophysics Data System (ADS)

Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

2016-11-01

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

DOE PAGES

Williamson, R. L.; Capps, N. A.; Liu, W.; ...

2016-09-27

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) ormore » plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used in this paper to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. Finally, in comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.« less

FUEL ELEMENT SUPPORT

DOEpatents

Wyman, W.L.

1961-06-27

The described cylindrical fuel element has longitudinally spaced sets of short longitudinal ribs circumferentially spaced from one another. The ribs support the fuel element in a coolant tube so that there is an annular space for coolant flow between the fuel element and the interior of the coolant tube. If the fuel element grows as a result of reactor operation, the circumferential distribution of the ribs maintains the uniformity of the annular space between the coolant tube and the fuel element, and the collapsibility of the ribs prevents the fuel element from becoming jammed in the coolant tube.

Tumorigenic Evaluation of Jet Fuels JP-TS and JP-7.

DTIC Science & Technology

1991-04-01

DTIC AL-TR-1991 0020 3 ELECTE0 AD-A252 012 JUN 2 6 1992• • TUMORIGENIC EVALUATION OF JET FUELS JP-TS AND JP-7 E. R. Kinkead C. L. Gaworski C. D...Evaluation of Jet Fuels JP-TS and JP-7. The research described in this report began in March 1981 and was completed in February 1991 under U.S. Air Force...of jet engines in military and commercial aircraft has led to the development of a number of petroleum distillate fuels with special properties. These

Thermal breeder fuel enrichment zoning

DOEpatents

Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

1992-01-01

A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

Co-Optimization of Fuels and Engines

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

Farrell, John

2016-03-24

The Co-Optimization of Fuels and Engines (Co-Optima) initiative is a new DOE initiative focused on accelerating the introduction of affordable, scalable, and sustainable biofuels and high-efficiency, low-emission vehicle engines. The simultaneous fuels and vehicles research and development (R&D) are designed to deliver maximum energy savings, emissions reduction, and on-road vehicle performance. The initiative's integrated approach combines the previously independent areas of biofuels and combustion R&D, bringing together two DOE Office of Energy Efficiency & Renewable Energy research offices, ten national laboratories, and numerous industry and academic partners to simultaneously tackle fuel and engine research and development (R&D) to maximize energymore » savings and on-road vehicle performance while dramatically reducing transportation-related petroleum consumption and greenhouse gas (GHG) emissions. This multi-year project will provide industry with the scientific underpinnings required to move new biofuels and advanced engine systems to market faster while identifying and addressing barriers to their commercialization. This project's ambitious, first-of-its-kind approach simultaneously tackles fuel and engine innovation to co-optimize performance of both elements and provide dramatic and rapid cuts in fuel use and emissions. This presentation provides an overview of the project.« less

2D and 3D Multiscale/Multicomponent Modeling of Impact Response of Heterogeneous Energetic Composites

DTIC Science & Technology

2016-06-01

7 Development of Cohesive Finite Element Method (CFEM) Capability ................................7 3D...Cohesive Finite Element Method (CFEM) framework A new scientific framework and technical capability is developed for the computational analyses of...this section should shift from reporting activities to reporting accomplishments. Development of Cohesive Finite Element Method (CFEM) Capability

A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2013-01-01

This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

Low temperature chemical processing of graphite-clad nuclear fuels

DOEpatents

Pierce, Robert A.

2017-10-17

A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

Characteristics of particulate emissions from a diesel generator fueled with varying blends of biodiesel and fossil diesel.

PubMed

Tsai, Jen-Hsiung; Chen, Shui-Jen; Huang, Kuo-Lin; Lee, Wen-Jhy; Kuo, Wen-Chien; Lin, Wen-Yinn

2011-01-01

This study investigated the particulate matter (PM), particle-bound carbons, and polycyclic aromatic hydrocarbons (PAHs) emitted from a diesel-engine generator fuelled with blends of pure fossil diesel oil (D100) and varying percentages of waste-edible-oil biodiesel (W10, 10 vol %; W20, 20 vol %; W30, 30 vol %; and W50, 50 vol %) under generator loads of 0, 1.5, and 3 kW. On average, the PM emission factors of all blends was 30.5 % (range, 13.7-52.3 %) lower than that of D100 under the tested loads. Substituting pure fossil diesel oil with varying percentages of waste-edible-oil biodiesel reduced emissions of particle-bound total carbon (TC) and elemental carbon (EC). The W20 blend had the lowest particle-bound organic carbon (OC) emissions. Notably, W10, W20, and W30 also had lower Total-PAH emissions and lower total equivalent toxicity (Total-BaP(eq)) compared to D100. Additionally, the brake-specific fuel consumption of the generator correlated positively with the ratio of waste-edible-oil biodiesel to pure fossil diesel. However, generator energy efficiency correlated negatively with the ratio of waste-edible-oil biodiesel to pure fossil diesel.

Fine woody fuel particle diameters for improved planar intersect fuel loading estimates in Southern Rocky Mountain ponderosa pine forests

Treesearch

Emma Vakili; Chad M. Hoffman; Robert E. Keane

2016-01-01

Fuel loading estimates from planar intersect sampling protocols for fine dead down woody surface fuels require an approximation of the mean squared diameter (d2) of 1-h (0-0.63 cm), 10-h (0.63-2.54 cm), and 100-h (2.54-7.62 cm) timelag size classes. The objective of this study is to determine d2 in ponderosa pine (Pinus ponderosa) forests of New Mexico and Colorado,...

77 FR 5418 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-03

... aft fuel system 40 micron fuel filter element with a 10 micron fuel filter element. This proposed AD... fuel filter element, part number (P/N) 52-0505-2 or 52-01064-1. This proposed AD would require replacing each forward and aft fuel system 40 micron fuel filter element with a 10 micron fuel filter...

40 CFR 80.1130 - Requirements for exporters of renewable fuels.

Code of Federal Regulations, 2010 CFR

2010-07-01

... gallons of renewable fuel. k = A discrete volume of renewable fuel. VOLk = The standardized volume of discrete volume k of exported renewable fuel, in gallons, calculated in accordance with § 80.1126(d)(7). EVk = The equivalence value associated with discrete volume k. = Sum involving all volumes of...

Design and evaluation of aircraft heat source systems for use with high-freezing point fuels

NASA Technical Reports Server (NTRS)

Pasion, A. J.

1979-01-01

The objectives were the design, performance and economic analyses of practical aircraft fuel heating systems that would permit the use of high freezing-point fuels on long-range aircraft. Two hypothetical hydrocarbon fuels with freezing points of -29 C and -18 C were used to represent the variation from current day jet fuels. A Boeing 747-200 with JT9D-7/7A engines was used as the baseline aircraft. A 9300 Km mission was used as the mission length from which the heat requirements to maintain the fuel above its freezing point was based.

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

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

Baek J.; Diamond D.; Cuadra, A.

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

Evaluation of Cetane Indices for Marine Fuels.

DTIC Science & Technology

1987-11-01

EIC. f I .ER TSO 7 WELLS AvEr UE . C0 E NEWTON CENTRE, NA 02159 K 0 *" 4’ OVEMBER 19 7. . DUh: "r w GEO-CENTERS, INC. Appvvv1d fo pbcmasuw...8217- ’ Just ifloatlio , ,S.- ......, . D istributloo/ K "vallablilty Codes Avail and/or AOlst Speoial j;;~ .. ; I j I ]J SI A ; A ( T 4%% ’-N4, N...established parameters as well as on "’ " . measurements such as refractive index and density: was GEO-CENTERS, INC. % %, K N 7A-, : The fuel set for

NUCLEAR REACTOR FUEL ELEMENT

DOEpatents

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

1961-09-01

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

Low cost, lightweight fuel cell elements

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor)

2001-01-01

New fuel cell elements for use in liquid feed fuel cells are provided. The elements including biplates and endplates are low in cost, light in weight, and allow high efficiency operation. Electrically conductive elements are also a part of the fuel cell elements.

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1958-08-19

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

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.

A Wind Tunnel Investigation to Determine Dominant Forebody Strake Design Characteristics for an F-15 Equipped with Conformal Fuel Tanks.

DTIC Science & Technology

1983-12-01

LONSlZTUDNAL STABILITY DATA FOR Nl F-15 WITH ONLY CFTJ! AND AMl F-13 WITH CFTS AND FIB STRAKES. ., 0-34 -... .4r * CL 1.4. .2 .2 .4 .6 .8 1 1.2 1.4 CO CM...CONFORMAL FUEL TANKS THESIS "AFIT/GAE/AA/83D-7 Terry A. DuncanCaptain USAF DT C SELECTE ca JAN 18 1984 DEPARTMENT OF THE AIR FORCE S AIR UNIVERSITY E AIR...DETERMINE DOMINANT FOREBODY STRAKE DESIGN CHARACTERISTICS FOR AN F-15 EQUIPPED WITH CONFORMAL FUEL TANKS THESIS AFITIGAE/AA/83D-7 Terry A. Duncan

RELAP5 Application to Accident Analysis of the NIST Research Reactor

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

Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.

Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accidentmore » and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.« less

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

DOEpatents

Honekamp, John R.; Fryer, Richard M.

1978-01-01

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

Examination of UC-ZrC after long term irradiation at thermionic temperature

NASA Technical Reports Server (NTRS)

Yang, L.; Johnson, H. O.

1972-01-01

Two fluoride tungsten clad UC-ZrC fueled capsules, designated as V-2C and V-2D, were examined a hot cell after irradiation in NASA Plum Brook Reactor at a maximum cladding temperature of 1930 K for 11,089 and 12,031 hours to burnups of 3.0 x 10 to the 20th power and 2.1 x 10 to the 20th power fission/c.c. respectively. Percentage of fission gas release from the fuel material was measured by radiochemical means. Cladding deformation, fuel-cladding interaction and microstructures of fuel, cladding, and fuel-cladding interface were studied metallographically. Compositions of dispersions in fuel, fuel matrix and fuel-cladding interaction layer were analyzed by electron microprobe techniques. Axial and radial distributions of burnup were determined by gamma-scan, autoradiography and isotopic burnup analysis. The results are presented and discussed in conjunction with the requirements of thermionic fuel elements for space power application.

IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

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

I. Glagolenko; D. Wachs; N. Woolstenhulme

2010-10-01

Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily duemore » to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.« less

Mobile electric power. Final report

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

Bloomfield, D.P.; Bloomfield, V.J.; Grosjean, P.D.

1995-12-01

The objective of this program was to develop a mobile fuel cell power supply for use by soldiers. The Century Series of 100 through 500 watt fuel cell power supplies was developed. The Century Series fuel cell power supplies are made up of a fuel cell stack, chemical hydride hydrogen supply, a fan and a controller. The FC-200, the 200 watt Century Series power supply, weighs 8.8 ib. and has a volume of 322 in.3. The operating point is 0.7 volt/cell at 125 ASF; a power density of 22.7 watts/lb. or 0.62 watts/in.3 and an energy density of 110 whr/lb.more » The prototype 750 whr hydrogen supply weighs 7 lbs. and has a volume of 193 in.3. The fuel elements weigh 0.45 lb. and require 0.79 lbs. of water. The FC-200 has powered a scooter requiring a starting current of three times the rated current of the stack. It has also powered a microclimate cooler. (KAR) P. 1.« less

Mobile electric power

NASA Astrophysics Data System (ADS)

Bloomfield, David P.; Bloomfield, Valerie J.; Grosjean, Paul D.; Kelland, James W.

1995-02-01

The objective of this program was to develop a mobile fuel cell power supply for use by soldiers. The Century Series of 100 through 500 watt fuel cell power supplies was developed. The Century Series fuel cell power supplies are made up of a fuel cell stack, chemical hydride hydrogen supply, a fan and a controller. The FC-200, the 200 watt Century Series power supply, weighs 8.8 lb. and has a volume of 322 cu in. The operating point is 0.7 volt/cell at 125 ASF; a power density of 22.7 watts/cu in. or 0.62 watts/cu in. and an energy density of 110 whr/lb. The prototype 750 whr hydrogen supply weighs 7 lbs. and has a volume of 193 cu in. The fuel elements weigh 0.45 lb. and require 0.79 lbs. of water. The FC-200 has powered a scooter requiring a starting current of three times the rated current of the stack. It has also powered a microclimate cooler.

Effect of Air Swirler Configuration on Lean Direct Injector Flow Structure and Combustion Performance with a 7-Point Lean Direct Injector Array

NASA Technical Reports Server (NTRS)

Hicks, Yolanda R.; Tacina, Kathleen M.; Anderson, Robert C.

2017-01-01

This paper examines the fundamentals of fuel-air mixing in a lean direct injection concept. Results are presented to investigate the effects of air swirler angle, element spacing, and center element offset on recirculation zone formation, flame stability and gaseous emissions.

40 CFR 90.7 - Reference materials.

Code of Federal Regulations, 2013 CFR

2013-07-01

... EMISSIONS FROM NONROAD SPARK-IGNITION ENGINES AT OR BELOW 19 KILOWATTS General § 90.7 Reference materials... Fuels by the Research Method Appendix A to subpart D, Table 3. ASTM D2700-92: Standard Test Method for...

40 CFR 90.7 - Reference materials.

Code of Federal Regulations, 2014 CFR

2014-07-01

... EMISSIONS FROM NONROAD SPARK-IGNITION ENGINES AT OR BELOW 19 KILOWATTS General § 90.7 Reference materials... Fuels by the Research Method Appendix A to subpart D, Table 3. ASTM D2700-92: Standard Test Method for...

40 CFR 90.7 - Reference materials.

Code of Federal Regulations, 2012 CFR

2012-07-01

... EMISSIONS FROM NONROAD SPARK-IGNITION ENGINES AT OR BELOW 19 KILOWATTS General § 90.7 Reference materials... Fuels by the Research Method Appendix A to subpart D, Table 3. ASTM D2700-92: Standard Test Method for...

A novel microbial fuel cell sensor with biocathode sensing element.

PubMed

Jiang, Yong; Liang, Peng; Liu, Panpan; Wang, Donglin; Miao, Bo; Huang, Xia

2017-08-15

The traditional microbial fuel cell (MFC) sensor with bioanode as sensing element delivers limited sensitivity to toxicity monitoring, restricted application to only anaerobic and organic rich water body, and increased potential fault warning to the combined shock of organic matter/toxicity. In this study, the biocathode for oxygen reduction reaction was employed for the first time as the sensing element in MFC sensor for toxicity monitoring. The results shown that the sensitivity of MFC sensor with biocathode sensing element (7.4±2.0 to 67.5±4.0mA% -1 cm -2 ) was much greater than that showed by bioanode sensing element (3.4±1.5 to 5.5±0.7mA% -1 cm -2 ). The biocathode sensing element achieved the lowest detection limit reported to date using MFC sensor for formaldehyde detection (0.0005%), while the bioanode was more applicable for higher concentration (>0.0025%). There was a quicker response of biocathode sensing element with the increase of conductivity and dissolved oxygen (DO). The biocathode sensing element made the MFC sensor directly applied to clean water body monitoring, e.g., drinking water and reclaimed water, without the amending of background organic matter, and it also decreased the warning failure when challenged by a combined shock of organic matter/toxicity. Copyright © 2017 Elsevier B.V. All rights reserved.

Fire Safety Tests Comparing Synthetic Jet and Diesel Fuels with JP-8 (POSTPRINT)

DTIC Science & Technology

2010-04-01

about 25% aromatics and 75% saturated (paraffin and naphthene ) hydro- carbons [5]. JP-8 is produced from jet fuel A by adding a corrosion inhibitor...4529a 43.9 44.2 Lubricity (mm) ASTM D – 5001 0.58 0.92 Acidity (mg KOH/g) ASTM D – 3242 0.004 0.003 SPK fuels taken from Moses [7], diesel fuels taken...this occurred the burnback pan was removed from the agent pan. The flames in the agent pan were allowed to break up the foam blanket and propagate until

Co-Optimization of Fuels and Engines

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

Farrell, John

2016-04-11

The Co-Optimization of Fuels and Engines (Co-Optima) initiative is a new DOE initiative focused on accelerating the introduction of affordable, scalable, and sustainable biofuels and high-efficiency, low-emission vehicle engines. The simultaneous fuels and vehicles research and development (R&D) are designed to deliver maximum energy savings, emissions reduction, and on-road vehicle performance. The initiative's integrated approach combines the previously independent areas of biofuels and combustion R&D, bringing together two DOE Office of Energy Efficiency & Renewable Energy research offices, ten national laboratories, and numerous industry and academic partners to simultaneously tackle fuel and engine research and development (R&D) to maximize energymore » savings and on-road vehicle performance while dramatically reducing transportation-related petroleum consumption and greenhouse gas (GHG) emissions. This multi-year project will provide industry with the scientific underpinnings required to move new biofuels and advanced engine systems to market faster while identifying and addressing barriers to their commercialization. This project's ambitious, first-of-its-kind approach simultaneously tackles fuel and engine innovation to co-optimize performance of both elements and provide dramatic and rapid cuts in fuel use and emissions. This presentation provides an overview of the initiative and reviews recent progress focused on both advanced spark-ignition and compression-ignition approaches.« less

CONCENTRIC TUBE FUEL ELEMENT SPRING ALIGNMENT SPACER DEVICE

DOEpatents

Weems, S.J.

1963-09-24

A rib construction for a nuclear-fuel element is described, in which one of three peripherally spaced ribs adjacent to each end of the fuel element is mounted on a radially yielding spring that embraces the fuel element. This spring enables the fuel element to have a good fit with a coolant tube and yet to be easily inserted in and withdrawn from the tube. (AEC)

Development of A 5,000 BBL, Rubberized Fabric Fuel Storage Tank, Collapsible,

DTIC Science & Technology

1981-04-01

Note l/ after soil burial. 6/ Reference fuel D is ASTM D-471, 60% iso-octane and 40% toluene. 7/ Retained after 56 days -25- i IGOODYEAR AEROSPACE 0 0...331.7) Pure. 9.I - 9.6 920 0() 09/O 0A) 2.7 ". paum Mese IEststa"t r.e ASIN 11-70 W1 I (ma) 7.3A -n-i I GAC 19-1337 Rev 2 USLE is (continmed) () The...the greater requirement. 5/ Method 5762 except that the specimens were prepared by Note 1/ after soil burial and the number of specimens was reduced

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

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

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

Utilization of D-beta-hydroxybutyrate and oleate as alternate energy fuels in brain cell cultures of newborn mice after hypoxia at different glucose concentrations.

PubMed

Bossi, E; Kohler, E; Herschkowitz, N

1989-11-01

In dissociated whole brain cell cultures from newborn mice, we have previously shown that during glucose deprivation under normoxia, D-beta-hydroxybutyrate and oleic acid are increasingly used for energy production. We now asked whether this glucose dependency of the utilization of D-beta-hydroxybutyrate and oleic acid as alternate energy fuels is also present after a hypoxic phase. 3-Hydroxy[3-14C]butyrate or [U-14C]oleic acid were added to 7- and 14-d-old cultures and 14CO2-production compared after hypoxia in normal and glucose-deprived conditions. After hypoxia, the ability of the cells 7 d in culture to increase D-beta-hydroxybutyrate consumption in response to glucose deprivation is diminished, 14-d-old cells lose this ability. In contrast, after hypoxia, both 7- and 14-d-old cultures maintain or even improve the ability to increase oleate consumption, when glucose is lacking.

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

DOEpatents

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

1962-08-14

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

15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. ...

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

15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. SHOWS AIR FORCE MAN AT EDGE OF TANK. INEL PHOTO NUMBER 65-6176, TAKEN NOVEMBER 10, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Transition Core Properties during Conversion of the NBSR from HEU to LEU Fuel

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

Hanson, A. L.; Diamond, D.

2013-10-31

The transition of the NBSR from HEU to LEU fuel is challenging due to reactivity constraints and the need to maintain an uninterrupted science program, the mission of the NBSR. The transition cannot occur with a full change of HEU to LEU fuel elements since the excess reactivity would be large enough that the NBSR would violate the technical specification for shutdown margin. Manufacturing LEU fuel elements to represent irradiated fuel elements would be cost prohibitive since 26 one-of-a-kind fuel elements would need to be manufactured. For this report a gradual transition from the present HEU fuel to the proposedmore » LEU fuel was studied. The gradual change approach would follow the present fuel management scheme and replace four HEU fuel elements with four LEU fuel elements each cycle. This manuscript reports the results of a series of calculations to predict the neutronic characteristics and how the neutronics will change during the transition from HEU to LEU in the NBSR.« less

NEUTRONIC REACTOR

DOEpatents

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

1958-09-01

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Credit WCT. Photographic copy of photograph, oxidizer and fuel tank ...

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

Credit WCT. Photographic copy of photograph, oxidizer and fuel tank assembly for engine tests being raised by crane for permanent installation in Test Stand "D" tower. Each tank held 170 gallons of propellants. (JPL negative 384-2029-B, 7 August 1959) - Jet Propulsion Laboratory Edwards Facility, Test Stand D, Edwards Air Force Base, Boron, Kern County, CA

Nuclear reactor

DOEpatents

Thomson, Wallace B.

2004-03-16

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

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.

Redwing: A MOOSE application for coupling MPACT and BISON

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

Frederick N. Gleicher; Michael Rose; Tom Downar

Fuel performance and whole core neutron transport programs are often used to analyze fuel behavior as it is depleted in a reactor. For fuel performance programs, internal models provide the local intra-pin power density, fast neutron flux, burnup, and fission rate density, which are needed for a fuel performance analysis. The fuel performance internal models have a number of limitations. These include effects on the intra-pin power distribution by nearby assembly elements, such as water channels and control rods, and the further limitation of applicability to a specified fuel type such as low enriched UO2. In addition, whole core neutronmore » transport codes need an accurate intra-pin temperature distribution in order to calculate neutron cross sections. Fuel performance simulations are able to model the intra-pin fuel displacement as the fuel expands and densifies. These displacements must be accurately modeled in order to capture the eventual mechanical contact of the fuel and the clad; the correct radial gap width is needed for an accurate calculation of the temperature distribution of the fuel rod. Redwing is a MOOSE-based application that enables coupling between MPACT and BISON for transport and fuel performance coupling. MPACT is a 3D neutron transport and reactor core simulator based on the method of characteristics (MOC). The development of MPACT began at the University of Michigan (UM) and now is under the joint development of ORNL and UM as part of the DOE CASL Simulation Hub. MPACT is able to model the effects of local assembly elements and is able calculate intra-pin quantities such as the local power density on a volumetric mesh for any fuel type. BISON is a fuel performance application of Multi-physics Object Oriented Simulation Environment (MOOSE), which is under development at Idaho National Laboratory. BISON is able to solve the nonlinearly coupled mechanical deformation and heat transfer finite element equations that model a fuel element as it is depleted in a nuclear reactor. Redwing couples BISON and MPACT in a single application. Redwing maps and transfers the individual intra-pin quantities such as fission rate density, power density, and fast neutron flux from the MPACT volumetric mesh to the individual BISON finite element meshes. For a two-way coupling Redwing maps and transfers the individual pin temperature field and axially dependent coolant densities from the BISON mesh to the MPACT volumetric mesh. Details of the mapping are given. Redwing advances the simulation with the MPACT solution for each depletion time step and then advances the multiple BISON simulations for fuel performance calculations. Sub-cycle advancement can be applied to the individual BISON simulations and allows multiple time steps to be applied to the fuel performance simulations. Currently, only loose coupling where data from a previous time step is applied to the current time step is performed.« less

Structural Evaluation of a Space Shuttle Main Engine (SSME) High Pressure Fuel Turbopump Turbine Blade

NASA Technical Reports Server (NTRS)

Abdul-Aziz, Ali

1996-01-01

Thermal and structural finite-element analyses were performed on the first high pressure fuel turbopump turbine blade of the space shuttle main engine (SSME). A two-dimensional (2-D) finite-element model of the blade and firtree disk attachment was analyzed using the general purpose MARC (finite-element) code. The loading history applied is a typical test stand engine cycle mission, which consists of a startup condition with two thermal spikes, a steady state and a shutdown transient. The blade material is a directionally solidified (DS) Mar-M 246 alloy, the blade rotor is forged with waspalloy material. Thermal responses under steady-state and transient conditions were calculated. The stresses and strains under the influence of mechanical and thermal loadings were also determined. The critical regions that exhibited high stresses and severe localized plastic deformation were the blade-rotor gaps.

Nuclear reactor control

DOEpatents

Cawley, William E.; Warnick, Robert F.

1982-01-01

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

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

Gleicher, Frederick; Ortensi, Javier; DeHart, Mark

Accurate calculation of desired quantities to predict fuel behavior requires the solution of interlinked equations representing different physics. Traditional fuels performance codes often rely on internal empirical models for the pin power density and a simplified boundary condition on the cladding edge. These simplifications are performed because of the difficulty of coupling applications or codes on differing domains and mapping the required data. To demonstrate an approach closer to first principles, the neutronics application Rattlesnake and the thermal hydraulics application RELAP-7 were coupled to the fuels performance application BISON under the master application MAMMOTH. A single fuel pin was modeledmore » based on the dimensions of a Westinghouse 17x17 fuel rod. The simulation consisted of a depletion period of 1343 days, roughly equal to three full operating cycles, followed by a station blackout (SBO) event. The fuel rod was depleted for 1343 days for a near constant total power loading of 65.81 kW. After 1343 days the fission power was reduced to zero (simulating a reactor shut-down). Decay heat calculations provided the time-varying energy source after this time. For this problem, Rattlesnake, BISON, and RELAP-7 are coupled under MAMMOTH in a split operator approach. Each system solves its physics on a separate mesh and, for RELAP-7 and BISON, on only a subset of the full problem domain. Rattlesnake solves the neutronics over the whole domain that includes the fuel, cladding, gaps, water, and top and bottom rod holders. Here BISON is applied to the fuel and cladding with a 2D axi-symmetric domain, and RELAP-7 is applied to the flow of the circular outer water channel with a set of 1D flow equations. The mesh on the Rattlesnake side can either be 3D (for low order transport) or 2D (for diffusion). BISON has a matching ring structure mesh for the fuel so both the power density and local burn up are copied accurately from Rattlesnake. At each depletion time step, Rattlesnake calculates a power density, fission density rate, burn-up distribution and fast flux based on the current water density and fuel temperature. These are then mapped to the BISON mesh for a fuels performance solve. BISON calculates the fuel temperature and cladding surface temperature based upon the current power density and bulk fluid temperature. RELAP-7 then calculates the fluid temperature, water density fraction and water phase velocity based upon the cladding surface temperature. The fuel temperature and the fluid density are then passed back to Rattlesnake for another neutronics calculation. Six Picard or fixed-point style iterations are preformed in this manner to obtain consistent tightly coupled and stable results. For this paper a set of results from the detailed calculation are provided for both during depletion and the SBO event. We demonstrate that a detailed calculation closer to first principles can be done under MAMMOTH between different applications on differing domains.« less

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Evidence for a single impact at the Cretaceous-Tertiary boundary from trace elements

NASA Technical Reports Server (NTRS)

Gilmour, Iain; Anders, Edward

1988-01-01

Not only meteoritic elements (Ir, Ni, Au, Pt metals), but also some patently non-meteoritic elements (As, Sb) are enriched at the K-T boundary. Eight enriched elements at 7 K-T sites were compared and it was found that: All have fairly constant proportions to Ir and Kilauea (invoked as an example of a volcanic source of Ir by opponents of the impact theory) has too little of 7 of these 8 elements to account for the boundary enrichments. The distribution of trace elements at the K-T boundary was reexamined using data from 11 sites for which comprehensive are available. The meteoritic component can be assessed by first normalizing the data to Ir, the most obviously extraterrestrial element, and then to Cl chondrites. The double normalization reduces the concentration range from 11 decades to 5 and also facilitates the identification of meteoritic elements. At sites where trace elements were analyzed in sub-divided samples of boundary clay, namely, Caravaca (SP), Stevns Klint (DK), Flaxbourne River (NZ) and Woodside Creek (NZ), Sb, As and Zn are well correlated with Ir across the boundary implying a common deposition mechanism. Elemental carbon is also enriched by up to 10,000 x in boundary clay from 5 K-T sides and is correlated with Ir across the boundary at Woodside Creek. While biomass would appear to be the primary fuel source for this carbon a contribution from a fossil fuel source may be necessary in order to account for the observed C abundance.

77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). DATES: Submit...

Conjugate Heat Transfer Analyses on the Manifold for Ramjet Fuel Injectors

NASA Technical Reports Server (NTRS)

Wang, Xiao-Yen J.

2006-01-01

Three-dimensional conjugate heat transfer analyses on the manifold located upstream of the ramjet fuel injector are performed using CFdesign, a finite-element computational fluid dynamics (CFD) software. The flow field of the hot fuel (JP-7) flowing through the manifold is simulated and the wall temperature of the manifold is computed. The three-dimensional numerical results of the fuel temperature are compared with those obtained using a one-dimensional analysis based on empirical equations, and they showed a good agreement. The numerical results revealed that it takes around 30 to 40 sec to reach the equilibrium where the fuel temperature has dropped about 3 F from the inlet to the exit of the manifold.

Co-Optimization of Fuels & Engines for Tomorrow's Energy-Efficient Vehicles

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

2016-03-01

A new U.S. Department of Energy (DOE) initiative is accelerating the introduction of affordable, scalable, and sustainable biofuels and high-efficiency, low-emission vehicle engines. The simultaneous fuels and vehicles research and development (R&D) is designed to deliver maximum energy savings, emissions reduction, and on-road vehicle performance. The initiative's integrated approach combines the previously independent areas of biofuels and combustion R&D, bringing together two DOE Office of Energy Efficiency & Renewable Energy research offices, nine national laboratories, and numerous industry and academic partners to more rapidly identify commercially viable solutions. This multi-year project will provide industry with the scientific underpinnings required tomore » move new biofuels and advanced engine systems to market faster while identifying and addressing barriers to their commercialization. This project's ambitious, first-of-its-kind approach simultaneously tackles fuel and engine innovation to co-optimize performance of both elements and provide dramatic and rapid cuts in fuel use and emissions.« less

DoD Mobility Fuels Guidelines Short-Term Issues,

DTIC Science & Technology

1979-06-01

built domestic plants . However, if no sub- stantial increase in domestic refinery capacity is forthcoming, this will encourage construction of new...from refinery investment. Under the Clean Air Act, refiners feel considerable uncertainty regarding fuel additives and refinery emissions . This has...II- 7 Market Structure ......... .................... II- 7 Refinery Capacity ........ .................... I- 9 Proposed Conversion to JP-8

Fuel pumping system and method

DOEpatents

Shafer, Scott F [Morton, IL; Wang, Lifeng ,

2006-12-19

A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

Fuel Pumping System And Method

DOEpatents

Shafer, Scott F.; Wang, Lifeng

2005-12-13

A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

Comparative Performance Assessment of 5kW-Class Solid Oxide Fuel Cell Engines Integrated With Single/Dual-Spool Turbochargers

DTIC Science & Technology

2011-01-01

Comparative Performance Assessment of 5kW-Class Solid Oxide Fuel Cell Engines Integrated with Single/Dual-Spool Turbochargers So-Ryeok Oh, Jing Sun... Turbochargers 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT... fundamental operating regime to the part load performance. Two different mechanical designs are assumed: dual shaft and single shaft as the compressor

7 CFR Appendix D to Subpart B of... - Technical Report for Flexible Fuel Pumps

Code of Federal Regulations, 2014 CFR

2014-01-01

... flexible fuel pump projects, as defined in § 4280.103. (a) Qualifications of project team. The flexible fuel pump project team is expected to consist of a project manager, an equipment supplier of major... may serve more than one role. Authoritative evidence that project team service providers have the...

7 CFR Appendix D to Subpart B of... - Technical Report for Flexible Fuel Pumps

Code of Federal Regulations, 2012 CFR

2012-01-01

... flexible fuel pump projects, as defined in § 4280.103. (a) Qualifications of project team. The flexible fuel pump project team is expected to consist of a project manager, an equipment supplier of major... may serve more than one role. Authoritative evidence that project team service providers have the...

7 CFR Appendix D to Subpart B of... - Technical Report for Flexible Fuel Pumps

Code of Federal Regulations, 2013 CFR

2013-01-01

... flexible fuel pump projects, as defined in § 4280.103. (a) Qualifications of project team. The flexible fuel pump project team is expected to consist of a project manager, an equipment supplier of major... may serve more than one role. Authoritative evidence that project team service providers have the...

dSAP18 and dHDAC1 contribute to the functional regulation of the Drosophila Fab-7 element.

PubMed

Canudas, Silvia; Pérez, Silvia; Fanti, Laura; Pimpinelli, Sergio; Singh, Navjot; Hanes, Steven D; Azorín, Fernando; Espinás, M Lluïsa

2005-01-01

It was described earlier that the Drosophila GAGA factor [Trithorax-like (Trl)] interacts with dSAP18, which, in mammals, was reported to be a component of the Sin3-HDAC co-repressor complex. GAGA-dSAP18 interaction was proposed to contribute to the functional regulation of the bithorax complex (BX-C). Here, we show that mutant alleles of Trl, dsap18 and drpd3/hdac1 enhance A6-to-A5 transformation indicating a contribution to the regulation of Abd-B expression at A6. In A6, expression of Abd-B is driven by the iab-6 enhancer, which is insulated from iab-7 by the Fab-7 element. Here, we report that GAGA, dSAP18 and dRPD3/HDAC1 co-localize to ectopic Fab-7 sites in polytene chromosomes and that mutant Trl, dsap18 and drpd3/hdac1 alleles affect Fab-7-dependent silencing. Consistent with these findings, chromatin immunoprecipitation analysis shows that, in Drosophila embryos, the endogenous Fab-7 element is hypoacetylated at histones H3 and H4. These results indicate a contribution of GAGA, dSAP18 and dRPD3/HDAC1 to the regulation of Fab-7 function.

The Nuclear Cryogenic Propulsion Stage

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Kim, Tony; Emrich, William J.; Hickman, Robert R.; Broadway, Jeramie W.; Gerrish, Harold P.; Belvin, Anthony D.; Borowski, Stanley K.; Scott, John H.

2014-01-01

Nuclear Thermal Propulsion (NTP) development efforts in the United States have demonstrated the technical viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes in a single burn (NRX-A6 test). Results from Project Rover indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. Excellent results were also obtained by the former Soviet Union. Although historical programs had promising results, many factors would affect the development of a 21st century nuclear thermal rocket (NTR). Test facilities built in the US during Project Rover no longer exist. However, advances in analytical techniques, the ability to utilize or adapt existing facilities and infrastructure, and the ability to develop a limited number of new test facilities may enable affordable development, qualification, and utilization of a Nuclear Cryogenic Propulsion Stage (NCPS). Bead-loaded graphite fuel was utilized throughout the Rover/NERVA program, and coated graphite composite fuel (tested in the Nuclear Furnace) and cermet fuel both show potential for even higher performance than that demonstrated in the Rover/NERVA engine tests.. NASA's NCPS project was initiated in October, 2011, with the goal of assessing the affordability and viability of an NCPS. FY 2014 activities are focused on fabrication and test (non-nuclear) of both coated graphite composite fuel elements and cermet fuel elements. Additional activities include developing a pre-conceptual design of the NCPS stage and evaluating affordable strategies for NCPS development, qualification, and utilization. NCPS stage designs are focused on supporting human Mars missions. The NCPS is being designed to readily integrate with the Space Launch System (SLS). A wide range of strategies for enabling affordable NCPS development, qualification, and utilization should be considered. These include multiple test and demonstration strategies (both ground and in-space), multiple potential test sites, and multiple engine designs. Two potential NCPS fuels are currently under consideration - coated graphite composite fuel and tungsten cermet fuel. During 2014 a representative, partial length (approximately 16") coated graphite composite fuel element with prototypic depleted uranium loading is being fabricated at Oak Ridge National Laboratory (ORNL). In addition, a representative, partial length (approximately 16") cermet fuel element with prototypic depleted uranium loading is being fabricated at Marshall Space Flight Center (MSFC). During the development process small samples (approximately 3" length) will be tested in the Compact Fuel Element Environmental Tester (CFEET) at high temperature (approximately 2800 K) in a hydrogen environment to help ensure that basic fuel design and manufacturing process are adequate and have been performed correctly. Once designs and processes have been developed, longer fuel element segments will be fabricated and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREE) at high temperature (approximately 2800 K) and in flowing hydrogen.

Means for supporting fuel elements in a nuclear reactor

DOEpatents

Andrews, Harry N.; Keller, Herbert W.

1980-01-01

A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

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.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Dickson, J.J.

1963-09-24

A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

40 CFR 90.7 - Reference materials.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Appendix A to subpart D, Table 3. ASTM D2699-92: Standard Test Method for Knock Characteristics of Motor... Knock Characteristics of Motor and Aviation Fuels by the Motor Method Appendix A to subpart D, Table 3...

40 CFR 90.7 - Reference materials.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Appendix A to subpart D, Table 3. ASTM D2699-92: Standard Test Method for Knock Characteristics of Motor... Knock Characteristics of Motor and Aviation Fuels by the Motor Method Appendix A to subpart D, Table 3...

Grooved Fuel Rings for Nuclear Thermal Rocket Engines

NASA Technical Reports Server (NTRS)

Emrich, William

2009-01-01

An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

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

West, W.S.

Progress during the period includes completion of the SNAP 7C system tests, completion of safety analysis for the SNAP 7A and C systems, assembly and initial testing of SNAP 7A, assembly of a modified reliability model, and assembly of a 10-W generator. Other activities include completion of thermal and safety analyses for SNAP 7B and D generators and fuel processing for these generators. (J.R.D.)

Short-Term Outlook for Hydrocarbon Gas Liquids

EIA Publications

2016-01-01

U.S. liquid fuels production increased from 7.43 million barrels per day (b/d) in 2008 to 13.75 million b/d in 2015. However, the Short-Term Energy Outlook (STEO) expects liquid fuels production to decline to 12.99 million b/d in 2017, mainly as a result of prolonged low oil prices. The liquid fuels production forecast reflects a 1.24 million b/d decline in crude oil production by 2017 that is partially offset by a 450,000 b/d increase in the production of hydrocarbon gas liquids (HGL)—a group of products including ethane, propane, butane (normal and isobutane), natural gasoline, and refinery olefins. This analysis will discuss the outlook for each of these four HGL streams and related infrastructure projects through 2017.

Current status of the development of high density LEU fuel for Russian research reactors

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

Vatulin, A.; Dobrikova, I.; Suprun, V.

2008-07-15

One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiationmore » examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)« less

Thermal-Hydraulic Transient Analysis of a Packed Particle Bed Reactor Fuel Element

DTIC Science & Technology

1990-06-01

long fuel elements, arranged to form a core , were analyzed for an up-power transient from 0 MWt to approximately 18 MWt. The simple model significantly...VARIATIONS IN FUEL ELEMENT GEOMETRY ............. 60 4.4 VARIATIONS IN THE MANNER OF TRANSIENT CONTROL ..... 62 4.5 CORE REPRESENTATION BY MULTIPLE FUEL ...the HTGR , however, the PBR packs small fuel particles between inner and outer retention elements, designated as frits. The PBR is appropriate for a

The use of graphene based materials for fuel cell, photovoltaics, and supercapacitor electrode materials

NASA Astrophysics Data System (ADS)

Tsang, Alpha C. H.; Kwok, Holly Y. H.; Leung, Dennis Y. C.

2017-05-01

This manuscript presents the methodology of the production of 2D and 3D graphene based material, and their applications in fuel cell, supercapacitor, and photovoltic in recent years. Due to the uniqueness and attractive properties of graphene nanosheets, a large number of techniques have been developed for raw graphene preparation, from a chemical method to a physical deposition of carbon vapor under extreme conditions. A variety of graphene based materials were also prepared from raw graphene or graphene oxide, including the metal loaded, metal oxides loaded, to the foreign elements doped graphene. Both two-dimensional (2D) to three-dimensional (3D) structured graphene were covered. These materials included the bulk or template hybrid composite, containing graphene hydrogel, graphene aerogel, or graphene foam and its derived products. They were widely used in green energy device research, which exhibited strong activity, and developed some special usage in recent research.

Nuclear fuel elements and method of making same

DOEpatents

Schweitzer, Donald G.

1992-01-01

A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

78 FR 17591 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-22

... aft fuel system 40 micron fuel filter element with a 10 micron nominal (40 micron absolute) fuel filter element. This AD was prompted by a National Transportation Safety Board (NTSB) review of in... helicopters with a fuel system 40 micron fuel filter element, part number (P/N) 52-0505-2 or 52-01064-1. That...

NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

DOEpatents

Szilard, L.; Young, G.J.

1958-03-01

This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

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

Montierth, Leland M.

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

Valorization of spent coffee grounds recycling as a potential alternative fuel resource in Turkey: An experimental study.

PubMed

Atabani, A E; Mercimek, S M; Arvindnarayan, Sundaram; Shobana, Sutha; Kumar, Gopalakrishnan; Cadir, Mehmet; Al-Muhatseb, Ala'a H

2018-03-01

In this study, recycling of spent coffee grounds (SCG) as a potential feedstock for alternative fuel production and compounds of added value in Turkey was assessed. The average oil content was found (≈ 13% w/w). All samples (before and after extraction) were tested for scanning electron microscopy (SEM), differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), X-ray diffraction (XRD), calorific value, surface analysis and porosity, Fourier transform infrared (FT-IR), and elemental analysis to assess their potential towards fuel properties. Elemental analysis indicated that carbon represents the highest percentages (49.59% and 46.42%, respectively), followed by nitrogen (16.7% and 15.5%), hydrogen (6.74% and 6.04%), and sulfur (0.851% and 0.561%). These results indicate that SCG can be utilized as compost, as it is rich in nitrogen. Properties of the extracted oil were examined, followed by biodiesel production. The quality of biodiesel was compared with American Society for Testing and Materials (ASTM) D6751 standards, and all the properties complied with standard specifications. The fatty acid compositions were analyzed by gas chromatography. It was observed that coffee waste methyl ester (CWME) is mainly composed of palmitic (35.8%) and arachidic (44.6%) acids, which are saturated fatty acids. The low degree of unsaturation provides an excellent oxidation stability (10.4 hr). CWME has also excellent cetane number, higher heating value, and iodine value with poor cold flow properties. The studies also investigated blending of biodiesel with Euro diesel and butanol. Following this, a remarkable improvement in cloud and pour points of biodiesel was obtained. Spent coffee grounds after oil extraction is an ideal material for garden fertilizer, feedstock for ethanol, biogas production, and as fuel pellets. The outcome of such research work produces valuable insights on the recycling importance of SCG in Turkey. Coffee is a huge industry, and coffee has been widely used due to its refreshing properties. This industry generates large quantities of waste. Therefore, recycling of spent coffee grounds for producing alternative fuels and compounds of added value is crucial. Elemental analysis indicated that coffee waste can be utilized as compost, as it is rich in nitrogen. Coffee waste after oil extraction is an ideal feedstock for ethanol and biogas production, garden fertilizer, and as fuel pellets. The low degree of unsaturation provides excellent oxidation stability. Its biodiesel has also excellent cetane number, higher heating value, and lower iodine value.

Inhalation exposure to JP-8 jet fuel alters pulmonary function and substance P levels in Fischer 344 rats.

PubMed

Pfaff, J; Parton, K; Lantz, R C; Chen, H; Hays, A M; Witten, M L

1995-01-01

In a simulated military flightline exposure protocol, Fischer 344 rats (F344) were used to investigate the pulmonary effects of JP-8 jet fuel inhalation. Exposures were nose only and for 1 h daily. Groups were exposed for 7 days (7D) or 28 days (28D). Each exposure group had a matched longitudinal control group (LC7 and LC28). Exposure concentrations of 520 mg m-3 caused an increase in dynamic compliance after 7 days of exposure, but compliance changes were not seen with continued exposure (28D, 495 mg m-3). Pulmonary resistance was increased in both 7- and 28-day JP-8-exposed groups. Changes in pulmonary function were accompanied by a decrease in substance P concentrations from the bronchoalveolar lavage fluid (BALF). No significant change was observed in BALF levels of 6-keto-PGF1 alpha, the stable metabolite of prostacyclin, which is a marker of endothelial cell function. The JP-8-exposed rats gained significantly less weight during the study period than the LC7 and LC28 groups, and the lungs of the 7D group were heavier by wet lung/body weight ratio (WtL/WtB). Alveolar clearance of technetium-labelled diethylenetriamine pentaacetate ([99mTc]DTPA) was increased in jet fuel-exposed groups. Light microscopy showed no pathological evidence of lung injury. Recovery from the early pulmonary effects of JP-8 inhalation occurred with continued exposure, as seen by recovery of pulmonary compliance and WtL/WtB.

Retrotransposons Are the Major Contributors to the Expansion of the Drosophila ananassae Muller F Element

PubMed Central

Shaffer, Christopher D.; Chen, Elizabeth J.; Quisenberry, Thomas J.; Ko, Kevin; Braverman, John M.; Giarla, Thomas C.; Mortimer, Nathan T.; Reed, Laura K.; Smith, Sheryl T.; Robic, Srebrenka; McCartha, Shannon R.; Perry, Danielle R.; Prescod, Lindsay M.; Sheppard, Zenyth A.; Saville, Ken J.; McClish, Allison; Morlock, Emily A.; Sochor, Victoria R.; Stanton, Brittney; Veysey-White, Isaac C.; Revie, Dennis; Jimenez, Luis A.; Palomino, Jennifer J.; Patao, Melissa D.; Patao, Shane M.; Himelblau, Edward T.; Campbell, Jaclyn D.; Hertz, Alexandra L.; McEvilly, Maddison F.; Wagner, Allison R.; Youngblom, James; Bedi, Baljit; Bettincourt, Jeffery; Duso, Erin; Her, Maiye; Hilton, William; House, Samantha; Karimi, Masud; Kumimoto, Kevin; Lee, Rebekah; Lopez, Darryl; Odisho, George; Prasad, Ricky; Robbins, Holly Lyn; Sandhu, Tanveer; Selfridge, Tracy; Tsukashima, Kara; Yosif, Hani; Kokan, Nighat P.; Britt, Latia; Zoellner, Alycia; Spana, Eric P.; Chlebina, Ben T.; Chong, Insun; Friedman, Harrison; Mammo, Danny A.; Ng, Chun L.; Nikam, Vinayak S.; Schwartz, Nicholas U.; Xu, Thomas Q.; Burg, Martin G.; Batten, Spencer M.; Corbeill, Lindsay M.; Enoch, Erica; Ensign, Jesse J.; Franks, Mary E.; Haiker, Breanna; Ingles, Judith A.; Kirkland, Lyndsay D.; Lorenz-Guertin, Joshua M.; Matthews, Jordan; Mittig, Cody M.; Monsma, Nicholaus; Olson, Katherine J.; Perez-Aragon, Guillermo; Ramic, Alen; Ramirez, Jordan R.; Scheiber, Christopher; Schneider, Patrick A.; Schultz, Devon E.; Simon, Matthew; Spencer, Eric; Wernette, Adam C.; Wykle, Maxine E.; Zavala-Arellano, Elizabeth; McDonald, Mitchell J.; Ostby, Kristine; Wendland, Peter; DiAngelo, Justin R.; Ceasrine, Alexis M.; Cox, Amanda H.; Docherty, James E.B.; Gingras, Robert M.; Grieb, Stephanie M.; Pavia, Michael J.; Personius, Casey L.; Polak, Grzegorz L.; Beach, Dale L.; Cerritos, Heaven L.; Horansky, Edward A.; Sharif, Karim A.; Moran, Ryan; Parrish, Susan; Bickford, Kirsten; Bland, Jennifer; Broussard, Juliana; Campbell, Kerry; Deibel, Katelynn E.; Forka, Richard; Lemke, Monika C.; Nelson, Marlee B.; O'Keeffe, Catherine; Ramey, S. Mariel; Schmidt, Luke; Villegas, Paola; Jones, Christopher J.; Christ, Stephanie L.; Mamari, Sami; Rinaldi, Adam S.; Stity, Ghazal; Hark, Amy T.; Scheuerman, Mark; Silver Key, S. Catherine; McRae, Briana D.; Haberman, Adam S.; Asinof, Sam; Carrington, Harriette; Drumm, Kelly; Embry, Terrance; McGuire, Richard; Miller-Foreman, Drew; Rosen, Stella; Safa, Nadia; Schultz, Darrin; Segal, Matt; Shevin, Yakov; Svoronos, Petros; Vuong, Tam; Skuse, Gary; Paetkau, Don W.; Bridgman, Rachael K.; Brown, Charlotte M.; Carroll, Alicia R.; Gifford, Francesca M.; Gillespie, Julie Beth; Herman, Susan E.; Holtcamp, Krystal L.; Host, Misha A.; Hussey, Gabrielle; Kramer, Danielle M.; Lawrence, Joan Q.; Martin, Madeline M.; Niemiec, Ellen N.; O'Reilly, Ashleigh P.; Pahl, Olivia A.; Quintana, Guadalupe; Rettie, Elizabeth A.S.; Richardson, Torie L.; Rodriguez, Arianne E.; Rodriguez, Mona O.; Schiraldi, Laura; Smith, Joanna J.; Sugrue, Kelsey F.; Suriano, Lindsey J.; Takach, Kaitlyn E.; Vasquez, Arielle M.; Velez, Ximena; Villafuerte, Elizabeth J.; Vives, Laura T.; Zellmer, Victoria R.; Hauke, Jeanette; Hauser, Charles R.; Barker, Karolyn; Cannon, Laurie; Parsamian, Perouza; Parsons, Samantha; Wichman, Zachariah; Bazinet, Christopher W.; Johnson, Diana E.; Bangura, Abubakarr; Black, Jordan A.; Chevee, Victoria; Einsteen, Sarah A.; Hilton, Sarah K.; Kollmer, Max; Nadendla, Rahul; Stamm, Joyce; Fafara-Thompson, Antoinette E.; Gygi, Amber M.; Ogawa, Emmy E.; Van Camp, Matt; Kocsisova, Zuzana; Leatherman, Judith L.; Modahl, Cassie M.; Rubin, Michael R.; Apiz-Saab, Susana S.; Arias-Mejias, Suzette M.; Carrion-Ortiz, Carlos F.; Claudio-Vazquez, Patricia N.; Espada-Green, Debbie M.; Feliciano-Camacho, Marium; Gonzalez-Bonilla, Karina M.; Taboas-Arroyo, Mariela; Vargas-Franco, Dorianmarie; Montañez-Gonzalez, Raquel; Perez-Otero, Joseph; Rivera-Burgos, Myrielis; Rivera-Rosario, Francisco J.; Eisler, Heather L.; Alexander, Jackie; Begley, Samatha K.; Gabbard, Deana; Allen, Robert J.; Aung, Wint Yan; Barshop, William D.; Boozalis, Amanda; Chu, Vanessa P.; Davis, Jeremy S.; Duggal, Ryan N.; Franklin, Robert; Gavinski, Katherine; Gebreyesus, Heran; Gong, Henry Z.; Greenstein, Rachel A.; Guo, Averill D.; Hanson, Casey; Homa, Kaitlin E.; Hsu, Simon C.; Huang, Yi; Huo, Lucy; Jacobs, Sarah; Jia, Sasha; Jung, Kyle L.; Wai-Chee Kong, Sarah; Kroll, Matthew R.; Lee, Brandon M.; Lee, Paul F.; Levine, Kevin M.; Li, Amy S.; Liu, Chengyu; Liu, Max Mian; Lousararian, Adam P.; Lowery, Peter B.; Mallya, Allyson P.; Marcus, Joseph E.; Ng, Patrick C.; Nguyen, Hien P.; Patel, Ruchik; Precht, Hashini; Rastogi, Suchita; Sarezky, Jonathan M.; Schefkind, Adam; Schultz, Michael B.; Shen, Delia; Skorupa, Tara; Spies, Nicholas C.; Stancu, Gabriel; Vivian Tsang, Hiu Man; Turski, Alice L.; Venkat, Rohit; Waldman, Leah E.; Wang, Kaidi; Wang, Tracy; Wei, Jeffrey W.; Wu, Dennis Y.; Xiong, David D.; Yu, Jack; Zhou, Karen; McNeil, Gerard P.; Fernandez, Robert W.; Menzies, Patrick Gomez; Gu, Tingting; Buhler, Jeremy; Mardis, Elaine R.; Elgin, Sarah C.R.

2017-01-01

The discordance between genome size and the complexity of eukaryotes can partly be attributed to differences in repeat density. The Muller F element (∼5.2 Mb) is the smallest chromosome in Drosophila melanogaster, but it is substantially larger (>18.7 Mb) in D. ananassae. To identify the major contributors to the expansion of the F element and to assess their impact, we improved the genome sequence and annotated the genes in a 1.4-Mb region of the D. ananassae F element, and a 1.7-Mb region from the D element for comparison. We find that transposons (particularly LTR and LINE retrotransposons) are major contributors to this expansion (78.6%), while Wolbachia sequences integrated into the D. ananassae genome are minor contributors (0.02%). Both D. melanogaster and D. ananassae F-element genes exhibit distinct characteristics compared to D-element genes (e.g., larger coding spans, larger introns, more coding exons, and lower codon bias), but these differences are exaggerated in D. ananassae. Compared to D. melanogaster, the codon bias observed in D. ananassae F-element genes can primarily be attributed to mutational biases instead of selection. The 5′ ends of F-element genes in both species are enriched in dimethylation of lysine 4 on histone 3 (H3K4me2), while the coding spans are enriched in H3K9me2. Despite differences in repeat density and gene characteristics, D. ananassae F-element genes show a similar range of expression levels compared to genes in euchromatic domains. This study improves our understanding of how transposons can affect genome size and how genes can function within highly repetitive domains. PMID:28667019

Low-Cost Jet Fuel Starter Design Study

DTIC Science & Technology

1974-12-02

2G 27 3^ 38 & 39 60 vi WflU I LIST OF TABLES (continued) TABLE NO, 7 D-l I>-2 TITLE PAGE NO, Sea Level Design Point Component...Improvements 60 Turbojet Performance Summary D-3 Turbofan Performance Summary D-5 vii 1 SECTION INTRODUCTION The purpose of this study was to define...temperature difference between the top and bot- tom of the starter, does not begin to have an effect until after 60 seconds from shutdown. The Jet fuel

FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Abbott, W.E.; Balent, R.

1958-09-16

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

Evaluation of the finite element fuel rod analysis code (FRANCO)

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

Lee, K.; Feltus, M.A.

1994-12-31

Knowledge of temperature distribution in a nuclear fuel rod is required to predict the behavior of fuel elements during operating conditions. The thermal and mechanical properties and performance characteristics are strongly dependent on the temperature, which can vary greatly inside the fuel rod. A detailed model of fuel rod behavior can be described by various numerical methods, including the finite element approach. The finite element method has been successfully used in many engineering applications, including nuclear piping and reactor component analysis. However, fuel pin analysis has traditionally been carried out with finite difference codes, with the exception of Electric Powermore » Research Institute`s FREY code, which was developed for mainframe execution. This report describes FRANCO, a finite element fuel rod analysis code capable of computing temperature disrtibution and mechanical deformation of a single light water reactor fuel rod.« less

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.

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

None

The objectives of the SNAP 7A program were to design, manufacture, test, and deliver a five-watt electric generation system for a U. S. Coast Guard 8 x 26E light buoy. The 10-watt Sr/sup 90/ thermoelectric generator, the d-c-to-d-c converter, batteries and the method of installation in the light buoy are describcd. The SNAP 7A generator was fueled with four capsules containing a total of 40,800 curies of Sr/sup 90/ titanate. After fueling and testing, the SNAP 7A electric generating system was installed in the Coast Guard light buoy at Baltimore, Maryland, on December 15, 1961. Operation of the buoy lampmore » is continuous. (auth)« less

Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Abdalla, Ayman

SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer Coefficient (HTC), bulk-fluid, sheath and fuel centerline temperature profiles were generated along the heated length of 5.772 m for a generic fuel channel. The fuel centerline and sheath temperature profiles have been determined at four Axial Heat Flux Profiles (AHFPs) using an average thermal power per channel of 8.5 MWth. The four examined AHFPs are the uniform, cosine, upstream-skewed and downstream-skewed profiles. Additionally, this study focuses on investigating a possibility of using low, enhanced and high thermal-conductivity fuels. The low thermal-conductivity fuels, which have been examined in this study, are uranium dioxide (UO 2), Mixed Oxide (MOX) and Thoria (ThO2) fuels. The examined enhanced thermal-conductivity fuels are uranium dioxide - silicon carbide (UO2 - SiC) and uranium dioxide - beryllium oxide (UO2 - BeO). Lastly, uranium carbide (UC), uranium dicarbide (UC2) and uranium nitride (UN) are the selected high thermal-conductivity fuels, which have been proposed for use in SCWRs. A comparison has been made between the low, enhanced and high thermal-conductivity fuels in order to identify the fuel centerline temperature behaviour when different nuclear fuels are used. Also, in the process of conducting the sensitivity analysis, the HTC was calculated using the Mokry et al. correlation, which is the most accurate supercritical water heat-transfer correlation so far. The sheath and the fuel centerline temperature profiles were determined for two cases. In Case 1, the HTC was calculated based on the Mokry et al. correlation, while in Case 2, the HTC values calculated for Case 1 were multiplied by a factor of 2. This factor was used in order to identify the amount of decrease in temperatures if the heat transfer is enhanced with appendages. Results of this analysis indicate that the use of the newly developed 54-element fuel bundle along with the proposed fuels is promising when compared with the Variant-20 (43-element) fuel bundle. Overall, the fuel centerline and sheath temperatures were below the industry and design limits when most of the proposed fuels were examined in the 54-element fuel bundle, however, the fuel centerline temperature limit was exceeded while MOX fuel was examined. Keywords: SCWRs, Fuel Centerline Temperature, Sheath Temperature, High Thermal Conductivity Fuels, Low Thermal Conductivity Fuels, HTC.

Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

Validating the BISON fuel performance code to integral LWR experiments

DOE PAGES

Williamson, R. L.; Gamble, K. A.; Perez, D. M.; ...

2016-03-24

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to datemore » for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

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)

Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

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

Eleon, Cyrille; Passard, Christian; Hupont, Nicolas

2015-07-01

Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no.more » 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)« less

Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

DTIC Science & Technology

1989-05-25

Engineer and Master of Science in Nuclear Engineering. ABSTRACT A model of the behavior of a packed bed nuclear reactor fuel element is developed . It...RECOMMENDATIONS FOR FURTHER INVESTIGATION .................... 150 APPENDIX A FUEL ELEMENT MODEL PROGRAM DESIGN AND OPERA- T IO N...follow describe the details of the packed bed reactor and then discuss the development of the mathematical representations of the fuel element. These are

METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY

DOEpatents

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

1961-06-27

A neutronic reactor comprising a moderator containing uniformly sized and spaced channels and uniformly dimensioned fuel elements is patented. The fuel elements have a fissionable core and an aluminum jacket. The cores and the jackets of the fuel elements in the central channels of the reactor are respectively thinner and thicker than the cores and jackets of the fuel elements in the remainder of the reactor, producing a flattened flux.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-16

A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

Spent nuclear fuel assembly inspection using neutron computed tomography

NASA Astrophysics Data System (ADS)

Pope, Chad Lee

The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

SLUG HANDLING DEVICES

DOEpatents

Gentry, J.R.

1958-09-16

A device is described for handling fuel elements of a neutronic reactor. The device consists of two concentric telescoped contalners that may fit about the fuel element. A number of ratchet members, equally spaced about the entrance to the containers, are pivoted on the inner container and spring biased to the outer container so thnt they are forced to hear against and hold the fuel element, the weight of which tends to force the ratchets tighter against the fuel element. The ratchets are released from their hold by raising the inner container relative to the outer memeber. This device reduces the radiation hazard to the personnel handling the fuel elements.

FUEL ELEMENTS FOR NUCLEAR REACTORS

DOEpatents

Blainey, A.; Lloyd, H.

1961-07-11

A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

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

Trammell, Michael P; Jolly, Brian C; Miller, James Henry

ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

Fuel element concept for long life high power nuclear reactors

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.; Rom, F. E.

1969-01-01

Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

Self Regulating Fiber Fuel Cell

DTIC Science & Technology

2010-08-16

12000 68.2 77.4 24/7 Extreme Rigid liquid hydrogen fuel cell Medis 68 X 97 X 57 20000 53.2 108.1 Fiber Fuel Cell Flexible Individual fiber Honeywell...which allows hydrogen and water vapor to permeate freely but prevents liquids from entering or fuel particles from escaping. The SPM permeability...S is the solubility and D is the diffusivity. Solubility and diffusivity data vs. pressure for hydrogen in Nafion is not available in the literature

3-D plasma boundary and plasma wall interaction research at UW-Madison

NASA Astrophysics Data System (ADS)

Schmitz, Oliver; Akerson, Adrian; Bader, Aaron; Barbui, Tullio; Effenberg, Florian; Flesch, Kurt; Frerichs, Heinke; Green, Jonathan; Hinson, Edward; Kremeyer, Thierry; Norval, Ryan; Stephey, Laurie; Waters, Ian; Winters, Victoria

2016-10-01

The necessity of considering 3-D effects on the plasma boundary and plasma wall interaction (PWI) in tokamaks, stellarators and reversed field pinches has been highlighted by abundant experimental and numerical results in the recent past. Prominent examples with 3-D boundary situations are numerous: ELM controlled H-modes by RMP fields in tokamaks, research on boundary plasmas and PWI in stellarators in general, quasi-helical states in RFPs, asymmetric fueling situations, and structural and wall elements which are not aligned with the magnetic guiding fields. A systematic approach is being taken at UW-Madison to establish a targeted experimental basis for identifying the most significant effects for plasma edge transport and resulting PWI in such 3-D plasma boundary situations. We deploy advanced 3-D modeling using the EMC3-EIRENE, ERO and MCI codes in combination with laboratory experiments at UW-Madison to investigate the relevance of 3-D effects in large scale devices with a concerted approach on DIII-D, NSTX-U, and Wendelstein 7-X. Highlights of experimental results from the on-site laboratory activities at UW-Madison and the large scale facilities are presented and interlinks will be discussed. This work was supported by US DOE DE-SC0013911, DE-SC00012315 and DE-SC00014210.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Shackleford, M.H.

1958-12-16

A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ...

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

35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ELEMENT HOLDER, TRIP MECHANISM COVER, AND OTHER DETAILS. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-3. INEL INDEX CODE NUMBER: 075 0701 60 851 151977. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Possible consequences of operation with KIVN fuel elements in K Zircaloy process tubes

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

Carlson, P.A.

1963-08-06

From considerations of the results of experimental simulations of non-axial placement of fuel elements in process tubes and in-reactor experience, it is concluded that the ultimate outcome of a charging error which results in operation with one or more unsupported fuel elements in a K Zircaloy-2 process tube would be multiple fuel failure and failure of the process tube. The outcome of the accident is determined by the speed with which the fuel failure is detected and the reactor is shut down. The release of fission products would be expected to be no greater than that which has occurred followingmore » severe fuel failure incidents. The highest probability for fission product release occurs during the discharge of failed fuel elements, when a small fraction of the exposed uranium of the fuel element may be oxidized when exposed to air before the element falls into the water-filled discharge chute. The confinement and fog spray facilities were installed to reduce the amount of fission products which might escape from the reactor building after such an event.« less

United States Air Force Graduate Student Summer Support Program (1987). Program Technical Report. Volume 1.

DTIC Science & Technology

1987-12-01

developed for a large percentage of the participants in the Summer Faculty Research Program in 1979-1983 period through an AFOSR Minigrant Program . On 1...Analysis of a Bimodal Nuclear Rocket Core by Dav,, C. Carpenter ABSTRACT The framework for a general purpose finite element analysis code was developed ...to study the 2-D temperature distribution in a hot-channel S hexagonal fuel element in the core of a bimodal nuclear’ rocket. Prelim- inary thermal

Proposal for a possible use of fusion power for hydrogen production within this century

NASA Astrophysics Data System (ADS)

Seifritz, W.

Consideration is given to the possibility of building a commercial fusion power reactor before the turn of the century. The main element incorporated by the proposed system is the PACER project powerplant, which employs the explosive deuterium-deuterium (D-D) fusion process. Because all required technology already exists, PACER is believed to represent the quickest way to harness fusion on a large scale. It is argued that such reactors, scattered throughout the world on a series of 'energy parks', will meet a 30 TW global energy demand after the depletion of fossil fuel resources. Consideration is also given to both the breeding of fissile materials and the electrolytic production of hydrogen; a by-product of which would be deuterium fuel.

Rack for storing spent nuclear fuel elements

DOEpatents

Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

1978-06-20

A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

METHOD OF OPERATING NUCLEAR REACTORS

DOEpatents

Untermyer, S.

1958-10-14

A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

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

Hursin, M.; Perret, G.

The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handlingmore » and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)« less

Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

DOEpatents

Sease, J.D.; Harrington, F.E.

1973-12-11

Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication

NASA Technical Reports Server (NTRS)

Mireles, O. R.; Broadway, J.; Hickman, R.

2014-01-01

Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.

Propulsion and Power Rapid Response Research and Development Support. Delivery Order 0042: Demonstration and Evaluation of Fischer-Tropsch Research Fuels for the DoD Assured Fuels Program

DTIC Science & Technology

2006-12-01

7 Figure 3.1-3. Net Heat of Combustion ...No. 1 Aromatic Carbon, max ASTM D-5292 Mol % < 0.5 Sulfur, Total, Max ASTM D-5453 ppm 1 Cetane Index ASTM D-976 Report Net Heat of Combustion , min ASTM...12 /2 00 6 8/ 19 /2 00 6 8/ 26 /2 00 6 9/ 2/ 20 06 8 Figure 3.1-3. Net Heat of Combustion Trend Figure 3.1-4. Freezing Point Trend Net

Preparation of high temperature gas-cooled reactor fuel element

DOEpatents

Bradley, Ronnie A.; Sease, John D.

1976-01-01

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

Nuclear fuel element

DOEpatents

Zocher, Roy W.

1991-01-01

A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

KINETICS OF TREAT USED AS A TEST REACTOR

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

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

Processing of U-2.5Zr-7.5Nb and U-3Zr-9Nb alloys by sintering process

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

Dos Santos, A. M. M.; Ferraz, W. B.; Lameiras, F. S.

2012-07-01

To minimize the risk of nuclear proliferation, there is worldwide interest in reducing fuel enrichment of research and test reactors. To achieve this objective while still guaranteeing criticality and cycle length requirements, there is need of developing high density uranium metallic fuels. Alloying elements such as Zr, Nb and Mo are added to uranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) is developing the U-2.5Zr-7.5Nb and U-3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materials in the form of powders. The powders were pressed atmore » 400 MPa and then sintered under a vacuum of about 1x10{sup -4} Torr at temperatures ranging from 1050 deg. to 1500 deg.C. The densities of the alloys were measured geometrically and by hydrostatic method and the phases identified by X ray diffraction (XRD). The microstructures of the pellets were observed by scanning electron microscopy (SEM) and the alloying elements were analyzed by energy dispersive X-ray spectroscopy (EDS). The results obtained showed the fuel density to slightly increase with the sintering temperature. The highest density achieved was approximately 80% of theoretical density. It was observed in the pellets a superficial oxide layer formed during the sintering process. (authors)« less

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

Konyashov, Vadim V.; Krasnov, Alexander M.

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

40 CFR 52.570 - Identification of plan.

Code of Federal Regulations, 2012 CFR

2012-07-01

.../99, 64 FR 67491 391-3-1-.02(2)(d) Fuel-burning Equipment 7/20/05 2/9/09, 75 FR 6309 391-3-1-.02(2)(e.../18/79, 44 FR 54047 391-3-1-.02(2)(k) Particulate Emission from Asphaltic Concrete Hot Mix Plants 1/17...) NOX Emissions from Fuel-burning Equipment 2/16/00 7/10/01, 66 FR 35906 391-3-1-.02(2)(mmm) NOX...

Engine component improvement: Performance improvement, JT9D-7 3.8 AR fan

NASA Technical Reports Server (NTRS)

Gaffin, W. O.

1980-01-01

A redesigned, fuel efficient fan for the JT9D-7 engine was tested. Tests were conducted to determine the effect of the 3.8 AR fan on performance, stability, operational characteristics, and noise of the JT9D-7 engine relative to the current 4.6 AR Bill-of-Material fan. The 3.8 AR fan provides increased fan efficiency due to a more advanced blade airfoil with increased chord, eliminating one part span shroud and reducing the number of fan blades and fan exit guide vanes. Engine testing at simulated cruise conditions demonstrated the predicted 1.3 percent improvement in specific fuel consumption with the redesigned 3.8 AR fan. Flight testing and sea level stand engine testing demonstrated exhaust gas temperature margins, fan and low pressure compressor stability, operational suitability, and noise levels comparable to the Bill-of-Material fan.

10 CFR Appendix D to Subpart D of... - Classes of Actions that Normally Require EISs

Code of Federal Regulations, 2014 CFR

2014-01-01

... [Reserved] D7 Contracts, policies, and marketing and allocation plans for electric power D8 Import or export... operational change D10 Treatment, storage, and disposal facilities for high-level waste and spent nuclear fuel... Contracts, Policies, and Marketing and Allocation Plans for Electric Power Establishment and implementation...

Photographic combustion characterization of LOX/Hydrocarbon type propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1980-01-01

One hundred twenty-seven tests were conducted over a chamber pressure range of 125-1500 psia, a fuel temperature range of -245 F to 158 F, and a fuel velocity range of 48-707 ft/sec to demonstrate the advantages and limitations of using high speed photography to identify potential combustion anomalies such as pops, fuel freezing, reactive stream separation and carbon formations. Combustion evaluation criteria were developed to guide selection of the fuels, injector elements, and operating conditions for testing. Separate criteria were developed for fuel and injector element selection and evaluation. The photographic test results indicated conclusively that injector element type and design directly influence carbon formation. Unlike spray fan, impingement elements reduce carbon formation because they induce a relatively rapid near zone fuel vaporization rate. Coherent jet impingement elements, on the other hand, exhibit increased carbon formation.

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

DOEpatents

Furgerson, W.T.

1963-12-17

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

NEUTRONIC REACTOR CHARGING AND DISCHARGING

DOEpatents

Zinn, W.H.

1959-07-14

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Fuel handling apparatus for a nuclear reactor

DOEpatents

Hawke, Basil C.

1987-01-01

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

Drying results of K-Basin fuel element 1990 (Run 1)

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

Marschman, S.C.; Abrefah, J.; Klinger, G.S.

1998-06-01

The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtainedmore » from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0.« less

FUEL ELEMENT FOR NEUTRONIC REACTORS

DOEpatents

Evans, T.C.; Beasley, E.G.

1961-01-17

A fuel element for neutronic reactors, particularly the gas-cooled type of reactor, is described. The element comprises a fuel-bearing plate rolled to form a cylinder having a spiral passageway passing from its periphery to its center. In operation a coolant is admitted to the passageway at the periphery of the element, is passed through the spiral passageway, and emerges into a central channel defined by the inner turn of the rolled plate. The advantage of the element is that the fully heated coolant (i.e., coolant emerging into the central channel) is separated and thus insulated from the periphery of the element, which may be in contact with a low-temperature moderator, by the intermediate turns of the spiral fuel element.

Beam heated linear theta-pinch device for producing hot plasmas

DOEpatents

Bohachevsky, Ihor O.

1981-01-01

A device for producing hot plasmas comprising a single turn theta-pinch coil, a fast discharge capacitor bank connected to the coil, a fuel element disposed along the center axis of the coil, a predetermined gas disposed within the theta-pinch coil, and a high power photon, electron or ion beam generator concentrically aligned to the theta-pinch coil. Discharge of the capacitor bank generates a cylindrical plasma sheath within the theta-pinch coil which heats the outer layer of the fuel element to form a fuel element plasma layer. The beam deposits energy in either the cylindrical plasma sheath or the fuel element plasma layer to assist the implosion of the fuel element to produce a hot plasma.

Dimension-six matrix elements for meson mixing and lifetimes from sum rules

NASA Astrophysics Data System (ADS)

Kirk, M.; Lenz, A.; Rauh, T.

2017-12-01

The hadronic matrix elements of dimension-six Δ F = 0, 2 operators are crucial inputs for the theory predictions of mixing observables and lifetime ratios in the B and D system. We determine them using HQET sum rules for three-point correlators. The results of the required three-loop computation of the correlators and the one-loop computation of the QCD-HQET matching are given in analytic form. For mixing matrix elements we find very good agreement with recent lattice results and comparable theoretical uncertainties. For lifetime matrix elements we present the first ever determination in the D meson sector and the first determination of Δ B = 0 matrix elements with uncertainties under control — superseeding preliminary lattice studies stemming from 2001 and earlier. With our state-of-the-art determination of the bag parameters we predict: τ( B +)/ τ( B d 0 ) = 1.082 - 0.026 + 0.022 , τ( B s 0 )/ τ( B d 0 ) = 0.9994 ± 0.0025, τ( D +)/ τ( D 0) = 2. 7 - 0.8 + 0.7 and the mixing-observables in the B s and B d system, in good agreement with the most recent experimental averages.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

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

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

Quantity and management of spent fuel from prototype and research reactors in Germany

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

Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on themore » information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)« less

BOILER-SUPERHEATED REACTOR

DOEpatents

Heckman, T.P.

1961-05-01

A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-11-21

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

Exclusion Area Radiation Release during the MIT Reactor Design Basis Accident.

DTIC Science & Technology

1983-05-06

Concrete Wall 116 6.2 Concrete Albedo Dose 121 6.3 Steel Door Scattering Dose 124 7.1 Total Dose Results 133 A.1 Values of N /NO for Neutron -Capture...plate fuel elements arranged in x a compact hexagonal core. This core design maximizes the neutron flux in the DO2 reflector region where numerous...sec) V = Volume of the fuel (cm 3 f Ef = Macroscopic fission cross section (cm ) = Thermal neutron flux ( neutrons /cm2 - sec) = Core-averaged value Yi

Survey of Advanced Propulsion Systems for Surface Vehicles

DTIC Science & Technology

1975-01-01

RROGRAM ELEMENT. PROJECT, T AM AREA • WORK UNIT NUMBERS T-102 12 . RERORT OATE January 1975 ’» NUMBER OF RAGES ZoL • S. SECURITY CLASS...wa*a«ilt,l,8>aaw^ •’ ..,-..,, I 1 I I 1 1 I I 0 a e 12 1 5 I 2. MILITARY NEEDS ~ Preceding page blank 21...sfc = endurance weight of fuel gross weight of vehicle power from engine specific fuel consumption 3-7-75- 12 40 60 80 100 120 140 SPECIFIC

Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

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

Pope, M. A.; DeHart, M. D.; Morrell, S. R.

2015-03-01

Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses,more » a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.« less

Physiology of the fuel ethanol strain Saccharomyces cerevisiae PE-2 at low pH indicates a context-dependent performance relevant for industrial applications.

PubMed

Della-Bianca, Bianca E; de Hulster, Erik; Pronk, Jack T; van Maris, Antonius J A; Gombert, Andreas K

2014-12-01

Selected Saccharomyces cerevisiae strains are used in Brazil to produce the hitherto most energetically efficient first-generation fuel ethanol. Although genome and some transcriptome data are available for some of these strains, quantitative physiological data are lacking. This study investigates the physiology of S. cerevisiae strain PE-2, widely used in the Brazilian fuel ethanol industry, in comparison with CEN.PK113-7D, a reference laboratory strain, focusing on tolerance to low pH and acetic acid stress. Both strains were grown in anaerobic bioreactors, operated as batch, chemostat or dynamic continuous cultures. Despite their different backgrounds, biomass and product formation by the two strains were similar under a range of conditions (pH 5 or pH < 3, with or without 105 mM acetic acid added). PE-2 displayed a remarkably higher fitness than CEN.PK113-7D during batch cultivation on complex Yeast extract - Peptone - Dextrose medium at low pH (2.7). Kinetics of viability loss of non-growing cells, incubated at pH 1.5, indicated a superior survival of glucose-depleted PE-2 cells, when compared with either CEN.PK113-7D or a commercial bakers' strain. These results indicate that the sulfuric acid washing step, used in the fuel ethanol industry to decrease bacterial contamination due to non-aseptic operation, might have exerted an important selective pressure on the microbial populations present in such environments. © 2014 Federation of European Microbiological Societies. Published by John Wiley & Sons Ltd. All rights reserved.

Bio-aviation fuel production from hydroprocessing castor oil promoted by the nickel-based bifunctional catalysts.

PubMed

Liu, Siyang; Zhu, Qingqing; Guan, Qingxin; He, Liangnian; Li, Wei

2015-05-01

Bio-aviation fuel was firstly synthesized by hydroprocessing castor oil in a continuous-flow fixed-bed microreactor with the main objective to obtain the high yield of aviation fuel and determine the elemental compositions of the product phases as well as the reaction mechanism. Highest aviation range alkane yields (91.6 wt%) were achieved with high isomer/n-alkane ratio (i/n) 4.4-7.2 over Ni supported on acidic zeolites. In addition, different fuel range alkanes can be obtained by adjusting the degree of hydrodeoxygenation (HDO) and hydrocracking. And the observations are rationalized by a set of reaction pathways for the various product phases. Copyright © 2015 Elsevier Ltd. All rights reserved.

Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

DOEpatents

Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

1982-01-01

A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

40 CFR 98.7 - What standardized methods are incorporated by reference into this part?

Code of Federal Regulations, 2010 CFR

2010-07-01

... Liquid Hydrocarbon Fuels by Bomb Calorimeter, IBR approved for §§ 98.34(a) and 98.254(e). (5) ASTM D388... Hydrocarbon Fuels by Bomb Calorimeter (Precision Method), IBR approved for §§ 98.34(a) and 98.254(e). (25...

40 CFR 52.380 - Rules and regulations.

Code of Federal Regulations, 2012 CFR

2012-07-01

... the following elements of the revisions: (1)-(2) [Reserved] (3) The program to review new and modified... (5) Stack height regulations (6) Interstate pollution requirements (7) Monitoring requirements (8) Conflict of interest provisions. (9) Use of 1 percent sulfur content fuel by the following residual oil...

40 CFR 52.380 - Rules and regulations.

Code of Federal Regulations, 2011 CFR

2011-07-01

... the following elements of the revisions: (1)-(2) [Reserved] (3) The program to review new and modified... (5) Stack height regulations (6) Interstate pollution requirements (7) Monitoring requirements (8) Conflict of interest provisions. (9) Use of 1 percent sulfur content fuel by the following residual oil...

40 CFR 52.380 - Rules and regulations.

Code of Federal Regulations, 2013 CFR

2013-07-01

... the following elements of the revisions: (1)-(2) [Reserved] (3) The program to review new and modified... (5) Stack height regulations (6) Interstate pollution requirements (7) Monitoring requirements (8) Conflict of interest provisions. (9) Use of 1 percent sulfur content fuel by the following residual oil...

40 CFR 52.380 - Rules and regulations.

Code of Federal Regulations, 2014 CFR

2014-07-01

... the following elements of the revisions: (1)-(2) [Reserved] (3) The program to review new and modified... (5) Stack height regulations (6) Interstate pollution requirements (7) Monitoring requirements (8) Conflict of interest provisions. (9) Use of 1 percent sulfur content fuel by the following residual oil...

40 CFR 52.570 - Identification of plan.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 391-3-1-.02(2)(c) Incinerators 6/15/98 12/2/99, 64 FR 67491 391-3-1-.02(2)(d) Fuel-burning Equipment 7...) Particulate Emission from Asphaltic Concrete Hot Mix Plants 1/17/79 9/18/79, 44 FR 54047 391-3-1-.02(2)(l.../2009 9/28/2012, 77 FR 59554 391-3-1-.02(2)(lll) NOX Emissions from Fuel-burning Equipment 2/16/00 7/10...

Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Guba, G. G.

2017-11-01

This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

Effect of High Si Content on U3Si2 Fuel Microstructure

NASA Astrophysics Data System (ADS)

Rosales, Jhonathan; van Rooyen, Isabella J.; Meher, Subhashish; Hoggan, Rita; Parga, Clemente; Harp, Jason

2018-02-01

The development of U3Si2 as an accident-tolerant nuclear fuel has gained research interest because of its promising high uranium density and improved thermal properties. In the present study, three samples of U3Si2 fuel with varying silicon content have been fabricated by a conventional powder metallurgical route. Microstructural characterization via scanning and transmission electron microscopy reveals the presence of other stoichiometry of uranium silicide such as USi and UO2 in both samples. The detailed phase analysis by x-ray diffraction shows the presence of secondary phases, such as USi, U3Si, and UO2. The samples with higher concentrations of silicon content of 7.5 wt.% display additional elemental Si. These samples also possess an increased amount of the USi phase as compared to that in the conventional sample with 7.3 wt.% silicon. The optimization of U3Si2 fuel performance through the understanding of the role of Si content on its microstructure has been discussed.

Using simulated 3D surface fuelbeds and terrestrial laser scan data to develop inputs to fire behavior models

Treesearch

Eric Rowell; E. Louise Loudermilk; Carl Seielstad; Joseph O' Brien

2016-01-01

Understanding fine-scale variability in understory fuels is increasingly important as physics-based fire behavior modelsdrive needs for higher-resolution data. Describing fuelbeds 3Dly is critical in determining vertical and horizontal distributions offuel elements and the mass, especially in frequently burned pine ecosystems where fine-scale...

Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

NASA Technical Reports Server (NTRS)

Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

2013-01-01

In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

Design of a fuel element for a lead-cooled fast reactor

NASA Astrophysics Data System (ADS)

Sobolev, V.; Malambu, E.; Abderrahim, H. Aït

2009-03-01

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg-1 of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg-1 of HM.

40 CFR 79.56 - Fuel and fuel additive grouping system.

Code of Federal Regulations, 2010 CFR

2010-07-01

... further testing under the provisions of Tier 3 or to support regulatory decisions affecting that fuel or... elements or classes of compounds other than those permitted in the base fuel for the respective fuel family... all of the following criteria: (1) Contain no elements other than carbon, hydrogen, oxygen, nitrogen...

Verification of radionuclide transfer factors to domestic-animal food products, using indigenous elements and with emphasis on iodine.

PubMed

Sheppard, S C; Long, J M; Sanipelli, B

2010-11-01

Recent reviews have established benchmark values for transfer factors that describe radionuclide transfer from plants to animal food product such as milk, eggs and meat. They also illustrate the paucity of data for some elements and some food products. The present study quantified transfer data using indigenous elements measured in dairy, poultry and other livestock farms in Canada. Up to 62 elements are reported, with particular emphasis on iodine (I) because of the need to accurately assess the behaviour of (129)I from disposal of nuclear fuel waste. There was remarkable agreement with the literature values, and for many elements the present study involved many more observations than were previously available. Perhaps the most important observation was that product/substrate concentration ratios (CR) were quite consistent across species, whereas the traditional fractional transfer factors (TF, units of d kg(-1) or d L(-1)) necessarily vary with body mass (feed intake). This suggests that for long-term assessments, it may be advisable to change the models to use CR rather than TF.

Advanced Thermally Stable Coal-Based Jet Fuels

DTIC Science & Technology

2007-10-01

has been minimized, additional dispersant does not affect deposition. Presumably, a second deposit 400 3PO Meat , Ntgen, S houn 3 JP8 N..t, AW, 5 h"Lrm...between 1.0 and 1.5. These fuel-rich equivalence ratios were 18 5 9 X-610 (RCO:LCO=1:1) A JP-8 4 JP-900 / /Tp= - 550K1 3 Ph = - 0.51 MPaMi*= 32 g/ s / 2...NUMBER F49620-99-1-0290 5C. PROGRAM ELEMENT NUMBER 61102F 6. AUTHOR( S ) 5d. PROJECT NUMBER 2308 Se. TASK NUMBER Harold H. Schobert BC 5f. WORK UNIT

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

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

Sell, D. A.; Baily, C. E.; Malewitz, T. J.

2016-09-01

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

Energy recycling by co-combustion of coal and recovered paint solids from automobile paint operations

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

Achariya Suriyawong; Rogan Magee; Ken Peebles

2009-05-15

This paper presents the results of an experimental study of particulate emission and the fate of 13 trace elements (arsenic (As), barium (Ba), cadmium (Cd), chromium (Cr), copper (Cu), cobalt (Co), manganese (Mn), molybdenum (Mo), nickel (Ni), lead (Pb), mercury (Hg), vanadium (V), and zinc (Zn)) during combustion tests of recovered paint solids (RPS) and coal. The emissions from combustions of coal or RPS alone were compared with those of co-combustion of RPS with subbituminous coal. The distribution/partitioning of these toxic elements between a coarse-mode ash (particle diameter (d{sub p}) > 0.5 {mu}m), a submicrometer-mode ash (d{sub p} < 0.5more » {mu}m), and flue gases was also evaluated. Submicrometer particles generated by combustion of RPS alone were lower in concentration and smaller in size than that from combustion of coal. However, co-combustion of RPS and coal increased the formation of submicrometer-sized particles because of the higher reducing environment in the vicinity of burning particles and the higher volatile chlorine species. Hg was completely volatilized in all cases; however, the fraction in the oxidized state increased with co-combustion. Most trace elements, except Zn, were retained in ash during combustion of RPS alone. Mo was mostly retained in all samples. The behavior of elements, except Mn and Mo, varied depending on the fuel samples. As, Ba, Cr, Co, Cu, and Pb were vaporized to a greater extent from cocombustion of RPS and coal than from combustion of either fuel. Evidence of the enrichment of certain toxic elements in submicrometer particles has also been observed for As, Cd, Cr, Cu, and Ni during co-combustion. 27 refs., 6 figs., 5 tabs.« less

Chlorine in solid fuels fired in pulverized fuel boilers sources, forms, reactions, and consequences: a literature review

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

David A. Tillman; Dao Duong; Bruce Miller

2009-07-15

Chlorine is a significant source of corrosion and deposition, both from coal and from biomass, and in PF boilers. This investigation was designed to highlight the potential for corrosion risks associated with once-through units and advanced cycles. The research took the form of a detailed literature investigation to evaluate chlorine in solid fuels: coals of various ranks and origins, biomass fuels of a variety of types, petroleum cokes, and blends of the above. The investigation focused upon an extensive literature review of documents dating back to 1991. The focus is strictly corrosion and deposition. To address the deposition and corrosionmore » issues, this review evaluates the following considerations: concentrations of chlorine in available solid fuels including various coals and biomass fuels, forms of chlorine in those fuels, and reactions - including reactivities - of chlorine in such fuels. The assessment includes consideration of alkali metals and alkali earth elements as they react with, and to, the chlorine and other elements (e.g., sulfur) in the fuel and in the gaseous products of combustion. The assessment also includes other factors of combustion: for example, combustion conditions including excess O{sub 2} and combustion temperatures. It also considers analyses conducted at all levels: theoretical calculations, bench scale laboratory data and experiments, pilot plant experiments, and full scale plant experience. Case studies and plant surveys form a significant consideration in this review. The result of this investigation focuses upon the concentrations of chlorine acceptable in coals burned exclusively, in coals burned with biomass, and in biomass cofired with coal. Values are posited based upon type of fuel and combustion technology. Values are also posited based upon both first principles and field experience. 86 refs., 8 figs., 7 tabs.« less

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.

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

Nuclear fuel elements having a composite cladding

DOEpatents

Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

1983-09-20

An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

40 CFR 52.570 - Identification of plan.

Code of Federal Regulations, 2013 CFR

2013-07-01

...) Incinerators 6/15/98 12/2/99, 64 FR 67491 391-3-1-.02(2)(d) Fuel-burning Equipment 7/20/05 2/9/09, 75 FR 6309... Concrete Hot Mix Plants 1/17/79 9/18/79, 44 FR 54047 391-3-1-.02(2)(l) Conical Burners 1/17/79 9/18/79, 44... Emissions from Fuel-burning Equipment 2/16/00 7/10/01, 66 FR 35906 391-3-1-.02(2)(mmm) NOX Emissions from...

An Experimental Study of Ignition Effects and Flame Growth Over a Thin Solid Fuel in Low-Speed Concurrent Flow Using Drop-Tower Facilities

NASA Technical Reports Server (NTRS)

Pettegrew, Richard Dale

1996-01-01

An experimental study of ignition and flame growth over a thin solid fuel in oxidizer flow speeds from 0 to 10 cm/sec concurrent flow was performed. This study examined the differences between ignition using a resistively heated wire (woven in a sawtooth pattern over the leading edge of the fuel), and a straight resistively heated wire augmented by a chemical ignitor doped onto the leading edge of the fuel. Results showed that the chemical system yielded non-uniform ignition bursts, while the system using only the hotwire gave more uniform ignition. At speeds up to 2.5 cm/sec, the chemical system yielded non-uniform pyrolysis fronts, while the hotwire system gave more uniform pyrolysis fronts. At speeds of 5 cm/sec or greater, both systems gave uniform pyrolysis fronts. The chemically-ignited flames tended to become too dim to see faster than the hotwire-ignited flames, and the flame lengths were observed to be shorter (after the initial burst subsided) for the chemical system for all speeds. Flame and pyrolysis element velocities were measured. Temperature profiles for selected tests were measured using thermocouples at the fuel surface and in the gas phase. Comparisons between the flame element velocities and peak temperatures recorded in these tests with calculated spread rates and peak temperatures from a steady-state model are presented. Agreement was found to be within 20% for most flame elements for nominal velocities of 5 cm/sec and 7.5 cm/sec.

Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.; Pearson, J. Boise; Schoenfeld, Michael P.

2015-01-01

The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at the time and resulted in NTREES being out of commission for a couple of months while a new stronger coil was procured. The new coil includes several additional pieces of support structure to prevent coil movement in the future. In addition, new insulating test article support components have been fabricated to prevent unexpected arcing to the test articles. Additional activities are also now underway to address ways in which the radial temperature profiles across test articles may be controlled such that they are more prototypical of what they would encounter in an operating nuclear engine. The causes of the temperature distribution problem are twofold. First, the fuel element test article is isolated in NTREES as opposed to being in the midst of many other mostly identical fuel elements in a nuclear engine. As a result, the fuel element heat flux boundary conditions in NTREES are far from adiabatic as would normally be the case in a reactor. Second, induction heating skews the power distribution such that power is preferentially deposited near the outside of the fuel element. Nuclear heating, conversely, deposits its power much more uniformly throughout the fuel element. Current studies are now looking at various schemes to adjust the amount of thermal radiation emitted from the fuel element surface so as to essentially vary the thermal boundary conditions on the test article. It is hoped that by properly adjusting the thermal boundary conditions on the fuel element test article, it may be possible to substantially correct for the inappropriate radial power distributions resulting from the induction heating so as to yield a more nearly correct temperature distribution throughout the fuel element.

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

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

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

METHOD AND APPARATUS FOR HANDLING RADIOACTIVE PRODUCTS

DOEpatents

Nicoll, D.

1959-02-24

A device is described for handling fuel elements being discharged from a nuclear reactor. The device is adapted to be disposed beneath a reactor within the storage canal for spent fuel elements. The device is comprised essentially of a cylinder pivotally mounted to a base for rotational motion between a vertical position. where the mouth of the cylinder is in the top portion of the container for receiving a fuel element discharged from a reactor into the cylinder, and a horizontal position where the mouth of the cylinder is remote from the top portion of the container and the fuel element is discharged from the cylinder into the storage canal. The device is operated by hydraulic pressure means and is provided with a means to prevent contaminated primary liquid coolant in the reactor system from entering the storage canal with the spent fuel element.

Calculation of Distribution Dynamics of Inhomogeneous Temperature Field in Range of Fuel Elements by Using FreeFem++

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Shishkin, A. V.

2017-11-01

This article introduces the result of studying the heat exchange in the fuel element of the nuclear reactor fuel magazine. Fuel assemblies are completed as a bundle of cylindrical fuel elements located at the tops of a regular triangle. Uneven distribution of fuel rods in a nuclear reactor’s core forms the inhomogeneity of temperature fields. This article describes the developed method for heat exchange calculation with the account for impact of an inhomogeneous temperature field on the thermal-physical properties of materials and unsteady effects. The acquired calculation results are used for evaluating the tolerable temperature levels in protective case materials.

Decomposition of the Seismic Source Using Numerical Simulations and Observations of Nuclear Explosions

DTIC Science & Technology

2017-05-31

SUBJECT TERMS nonlinear finite element calculations, nuclear explosion monitoring, topography 16. SECURITY CLASSIFICATION OF: 17. LIMITATION OF ABSTRACT 18...3D North Korea calculations........ Figure 6. The CRAM 3D finite element outer grid (left) is rectangular......................... Figure 7. Stress...Figure 6. The CRAM 3D finite element outer grid (left) is rectangular. The inner grid (center) is shaped to match the shape of the explosion shock wave

Analysis of resonance response performance of C-band antenna using parasitic element.

PubMed

Zaman, M R; Islam, M T; Misran, N; Mandeep, J S

2014-01-01

Analysis of the resonance response improvement of a planar C-band (4-8 GHz) antenna is proposed using parasitic element method. This parasitic element based method is validated for change in the active and parasitic antenna elements. A novel dual-band antenna for C-band application covering 5.7 GHz and 7.6 GHz is designed and fabricated. The antenna is composed of circular parasitic element with unequal microstrip lines at both sides and a rectangular partial ground plane. A fractional bandwidth of 13.5% has been achieved from 5.5 GHz to 6.3 GHz (WLAN band) for the lower band. The upper band covers from 7.1 GHz to 8 GHz with a fractional bandwidth of 12%. A gain of 6.4 dBi is achieved at the lower frequency and 4 dBi is achieved at the upper frequency. The VSWR of the antenna is less than 2 at the resonance frequency.

Fuel cell elements with improved water handling capacity

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor); Lee, Albany (Inventor)

2001-01-01

New fuel cell components for use in liquid feed fuel cell systems are provided. The components include biplates and endplates, having a hydrophilic surface and allow high efficiency operation. Conductive elements and a wicking device also form a part of the fuel cell components of the invention.

40 CFR 80.45 - Complex emissions model.

Code of Federal Regulations, 2010 CFR

2010-07-01

... this section for the appropriate phase and season Nd = exp d1(t)/exp d1(b) Hd = exp d2(t)/exp d2(b) w1.... d2(t) = Higher emitter 1,3-butadiene equation as defined in paragraph (e)(7)(ii) of this section... the base fuel's properties. d2(b) = Higher emitter 1,3-butadiene equation as defined in paragraph (e...

Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

NASA Astrophysics Data System (ADS)

Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

2017-04-01

Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

CFD Evaluation of a 3rd Generation LDI Combustor

NASA Technical Reports Server (NTRS)

Ajmani, Kumud; Mongia, Hukam; Lee, Phil

2017-01-01

An effort was undertaken to perform CFD analysis of fluid flow in Lean-Direct Injection (LDI) combustors with axial swirl-venturi elements for next-generation LDI-3 combustor design. The National Combustion Code (NCC) was used to perform non-reacting and two-phase reacting flow computations for a nineteen-element injector array arranged in a three-module, 7-5-7 element configuration. All computations were performed with a consistent approach of mesh-optimization, spray-modeling, ignition and kinetics-modeling with the NCC. Computational predictions of the aerodynamics of the injector were used to arrive at an optimal injector design that meets effective area and fuel-air mixing criteria. LDI-3 emissions (EINOx, EICO and UHC) were compared with the previous generation LDI-2 combustor experimental data at representative engine cycle conditions.

Method of forming a package for MEMS-based fuel cell

DOEpatents

Morse, Jeffrey D; Jankowski, Alan F

2013-05-21

A MEMS-based fuel cell package and method thereof is disclosed. The fuel cell package comprises seven layers: (1) a sub-package fuel reservoir interface layer, (2) an anode manifold support layer, (3) a fuel/anode manifold and resistive heater layer, (4) a Thick Film Microporous Flow Host Structure layer containing a fuel cell, (5) an air manifold layer, (6) a cathode manifold support structure layer, and (7) a cap. Fuel cell packages with more than one fuel cell are formed by positioning stacks of these layers in series and/or parallel. The fuel cell package materials such as a molded plastic or a ceramic green tape material can be patterned, aligned and stacked to form three dimensional microfluidic channels that provide electrical feedthroughs from various layers which are bonded together and mechanically support a MEMS-based miniature fuel cell. The package incorporates resistive heating elements to control the temperature of the fuel cell stack. The package is fired to form a bond between the layers and one or more microporous flow host structures containing fuel cells are inserted within the Thick Film Microporous Flow Host Structure layer of the package.

Method of forming a package for mems-based fuel cell

DOEpatents

Morse, Jeffrey D.; Jankowski, Alan F.

2004-11-23

A MEMS-based fuel cell package and method thereof is disclosed. The fuel cell package comprises seven layers: (1) a sub-package fuel reservoir interface layer, (2) an anode manifold support layer, (3) a fuel/anode manifold and resistive heater layer, (4) a Thick Film Microporous Flow Host Structure layer containing a fuel cell, (5) an air manifold layer, (6) a cathode manifold support structure layer, and (7) a cap. Fuel cell packages with more than one fuel cell are formed by positioning stacks of these layers in series and/or parallel. The fuel cell package materials such as a molded plastic or a ceramic green tape material can be patterned, aligned and stacked to form three dimensional microfluidic channels that provide electrical feedthroughs from various layers which are bonded together and mechanically support a MEMOS-based miniature fuel cell. The package incorporates resistive heating elements to control the temperature of the fuel cell stack. The package is fired to form a bond between the layers and one or more microporous flow host structures containing fuel cells are inserted within the Thick Film Microporous Flow Host Structure layer of the package.

78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Research and Test Reactors.'' This guide describes a method that the staff of the NRC considers acceptable... assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research...

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.

1957-10-01

A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

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

5 CFR 430.206 - Planning performance.

Code of Federal Regulations, 2013 CFR

2013-01-01

... summary level, including at least one critical element and any non-critical element(s). (5) Each... summary levels (pattern A as specified in § 430.208(d)(1)) shall not include non-critical elements. (7) An... non-critical elements. (8) Elements and standards shall be established as follows— (i) For a critical...

Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2017-01-01

To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

Space shuttle orbit maneuvering engine, reusable thrust chamber program. Task 6: Data dump hot fuel element investigation

NASA Technical Reports Server (NTRS)

Nurick, W. H.

1974-01-01

An evaluation of reusable thrust chambers for the space shuttle orbit maneuvering engine was conducted. Tests were conducted using subscale injector hot-fire procedures for the injector configurations designed for a regenerative cooled engine. The effect of operating conditions and fuel temperature on combustion chamber performance was determined. Specific objectives of the evaluation were to examine the optimum like-doublet element geometry for operation at conditions consistent with a fuel regeneratively cooled engine (hot fuel, 200 to 250 F) and the sensitivity of the triplet injector element to hot fuels.

Tag gas capsule with magnetic piercing device

DOEpatents

Nelson, Ira V.

1976-06-22

An apparatus for introducing a tag (i.e., identifying) gas into a tubular nuclear fuel element. A sealed capsule containing the tag gas is placed in the plenum in the fuel tube between the fuel and the end cap. A ferromagnetic punch having a penetrating point is slidably mounted in the plenum. By external electro-magnets, the punch may be caused to penetrate a thin rupturable end wall of the capsule and release the tag gas into the fuel element. Preferably the punch is slidably mounted within the capsule, which is in turn loaded as a sealed unit into the fuel element.

Nuclear reactor fuel element having improved heat transfer

DOEpatents

Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

1982-03-03

A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

Fuels Combustion Research.

DTIC Science & Technology

1982-07-30

Institute, Paper 78-42 (1978). 3. Colket, M.B., Naegeli , D.W. and Glassman, I., "High Temperature Pyro- lysis of Acetaldehyde", Int’l J. of Chem...Kinetics 7. 233 (1975). 4. Aronowitz, D. and Naegeli , D.W., "High Temperature Pyrolysis of Di-methylether", Int’l J. Chem. Kinetics_ 471 (1977). • 5

27 CFR 555.202 - Classes of explosive materials.

Code of Federal Regulations, 2010 CFR

2010-04-01

..., except for bulk salutes). (c) Blasting agents. (For example, ammonium nitrate-fuel oil and certain water-gels (see also § 555.11). [T.D. ATF-87, 46 FR 40384, Aug. 7, 1981, as amended by T.D. ATF-293, 55 FR 3722, Feb. 5, 1990; T.D. ATF-400, 63 FR 45003, Aug. 24, 1998] ...

Evaluation of Thin Plate Hydrodynamic Stability through a Combined Numerical Modeling and Experimental Effort

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

Tentner, A.; Bojanowski, C.; Feldman, E.

An experimental and computational effort was undertaken in order to evaluate the capability of the fluid-structure interaction (FSI) simulation tools to describe the deflection of a Missouri University Research Reactor (MURR) fuel element plate redesigned for conversion to lowenriched uranium (LEU) fuel due to hydrodynamic forces. Experiments involving both flat plates and curved plates were conducted in a water flow test loop located at the University of Missouri (MU), at conditions and geometries that can be related to the MURR LEU fuel element. A wider channel gap on one side of the test plate, and a narrower on the othermore » represent the differences that could be encountered in a MURR element due to allowed fabrication variability. The difference in the channel gaps leads to a pressure differential across the plate, leading to plate deflection. The induced plate deflection the pressure difference induces in the plate was measured at specified locations using a laser measurement technique. High fidelity 3-D simulations of the experiments were performed at MU using the computational fluid dynamics code STAR-CCM+ coupled with the structural mechanics code ABAQUS. Independent simulations of the experiments were performed at Argonne National Laboratory (ANL) using the STAR-CCM+ code and its built-in structural mechanics solver. The simulation results obtained at MU and ANL were compared with the corresponding measured plate deflections.« less

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

DOEpatents

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

1957-11-12

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

MRT fuel element inspection at Dounreay

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

Gibson, J.

1997-08-01

To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

Effect of Intake Pressure and Temperature on Auto-Ignition of Fuels with Different Cetane Number and Volatility

DTIC Science & Technology

2012-04-12

1 Effect Of Intake Pressure And Temperature On Auto-Ignition Of Fuels With Different Cetane Number And Volatility C. Jayakumar , U. Joshi, Z...AUTHOR(S) Eric Sattler; C. Jayakumar ; U. Joshi; Z. Zheng; W. Bryzik 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER 7. PERFORMING

Direct carbon fuel cell and stack designs

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

Gorte, Raymond J.; Oh, Tae-Sik

Disclosed are novel configurations of Direct Carbon Fuel Cells (DCFCs), which optionally comprise a liquid anode. The liquid anode comprises a molten salt/metal, preferably Sb, and a fuel, which has significant elemental carbon content (coal, bio-mass, etc.). The supply of fuel is continuously replenished in the anode. In addition, a stack configuration is suggested where combining a large number of planar or tubular fuel elements.

Nuclear breeder reactor fuel element with axial tandem stacking and getter

DOEpatents

Gibby, Ronald L.; Lawrence, Leo A.; Woodley, Robert E.; Wilson, Charles N.; Weber, Edward T.; Johnson, Carl E.

1981-01-01

A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

Construction of combustion models for rapeseed methyl ester bio-diesel fuel for internal combustion engine applications.

PubMed

Golovitchev, Valeri I; Yang, Junfeng

2009-01-01

Bio-diesel fuels are non-petroleum-based diesel fuels consisting of long chain alkyl esters produced by the transesterification of vegetable oils, that are intended for use (neat or blended with conventional fuels) in unmodified diesel engines. There have been few reports of studies proposing theoretical models for bio-diesel combustion simulations. In this study, we developed combustion models based on ones developed previously. We compiled the liquid fuel properties, and the existing detailed mechanism of methyl butanoate ester (MB, C(5)H(10)O(2)) oxidation was supplemented by sub-mechanisms for two proposed fuel constituent components, C(7)H(16) and C(7)H(8)O (and then, by mp2d, C(4)H(6)O(2) and propyne, C(3)H(4)) to represent the combustion model for rapeseed methyl ester described by the chemical formula, C(19)H(34)O(2) (or C(19)H(36)O(2)). The main fuel vapor thermal properties were taken as those of methyl palmitate C(19)H(36)O(2) in the NASA polynomial form of the Burcat database. The special global reaction was introduced to "crack" the main fuel into its constituent components. This general reaction included 309 species and 1472 reactions, including soot and NO(x) formation processes. The detailed combustion mechanism was validated using shock-tube ignition-delay data under diesel engine conditions. For constant volume and diesel engine (Volvo D12C) combustion modeling, this mechanism could be reduced to 88 species participating in 363 reactions.

Two-Dimensional Diffusion Theory Analysis of Reactivity Effects of a Fuel-Plate-Removal Experiment

NASA Technical Reports Server (NTRS)

Gotsky, Edward R.; Cusick, James P.; Bogart, Donald

1959-01-01

Two-dimensional two-group diffusion calculations were performed on the NASA reactor simulator in order to evaluate the reactivity effects of fuel plates removed successively from the center experimental fuel element of a seven- by three-element core loading at the Oak Ridge Bulk Shielding Facility. The reactivity calculations were performed by two methods: In the first, the slowing-down properties of the experimental fuel element were represented by its infinite media parameters; and, in the second, the finite size of the experimental fuel element was recognized, and the slowing-down properties of the surrounding core were attributed to this small region. The latter calculation method agreed very well with the experimented reactivity effects; the former method underestimated the experimental reactivity effects.

The General Formulation and Practical Calculation of the Diffusion Coefficient in a Lattice Containing Cavities; FORMULATION GENERALE ET CALCUL PRATIQUE DU COEFFICIENT DE DIFFUSION DANS UN RESEAU COMPORTANT DES CAVITES (in French)

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

Benoist, P.

The calculation of diffusion coefficients in a lattice necessitates the knowledge of a correct method of weighting the free paths of the different constituents. An unambiguous definition of this weighting method is given here, based on the calculation of leakages from a zone of a reactor. The formulation obtained, which is both simple and general, reduces the calculation of diffusion coefficients to that of collision probabilities in the different media; it reveals in the expression for the radial coefficient the series of the terms of angular correlation (cross terms) recently shown by several authors. This formulation is then used tomore » calculate the practical case of a classical type of lattice composed of a moderator and a fuel element surrounded by an empty space. Analytical and numerical comparison of the expressions obtained with those inferred from the theory of BEHRENS shows up the importance of several new terms some of which are linked with the transparency of the fuel element. Cross terms up to the second order are evaluated. A practical formulary is given at the end of the paper. (author) [French] Le calcul des coefficients de diffusion dans un reseau suppose la connaissance d'un mode de ponderation correct des libres parcours des differents constituants. On definit ici sans ambiguite ce mode de ponderation a partir du calcul des fuites hors d'une zone de reacteur. La formulation obtenue, simple et generale, ramene le calcul des coefficients de diffusion a celui des probabilites de collision dans les differents milieux; elle fait apparaitre dans l'expression du coefficient radial la serie des termes de correlation angulaire (termes rectangles), mis en evidence recemment par plusieurs auteurs. Cette formulation est ensuite appliquee au calcul pratique d'un reseau classique, compose d'un moderateur et d'un element combustible entoure d'une cavite; la comparaison analytique et numerique des expressions obtenues avec celles deduites de la theorie de BEHRENS fait apparaitre l'importance de plusieurs termes nouveaux, dont certains sont lies a la transparence de l'element combustible; les termes rectangles sont calcules jusqu'a l'ordre 2. Un formulaire pratique est donne a la fin de cette etude. (auteur)« less

Leaf-architectured 3D Hierarchical Artificial Photosynthetic System of Perovskite Titanates Towards CO2 Photoreduction Into Hydrocarbon Fuels

PubMed Central

Zhou, Han; Guo, Jianjun; Li, Peng; Fan, Tongxiang; Zhang, Di; Ye, Jinhua

2013-01-01

The development of an “artificial photosynthetic system” (APS) having both the analogous important structural elements and reaction features of photosynthesis to achieve solar-driven water splitting and CO2 reduction is highly challenging. Here, we demonstrate a design strategy for a promising 3D APS architecture as an efficient mass flow/light harvesting network relying on the morphological replacement of a concept prototype-leaf's 3D architecture into perovskite titanates for CO2 photoreduction into hydrocarbon fuels (CO and CH4). The process uses artificial sunlight as the energy source, water as an electron donor and CO2 as the carbon source, mimicking what real leaves do. To our knowledge this is the first example utilizing biological systems as “architecture-directing agents” for APS towards CO2 photoreduction, which hints at a more general principle for APS architectures with a great variety of optimized biological geometries. This research would have great significance for the potential realization of global carbon neutral cycle. PMID:23588925

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

Experimental Studies of Diestrol-Micro Emulsion Fuel in a Direct Injection Compression Ignition Engine under Varying Injection Pressures and Timings

NASA Astrophysics Data System (ADS)

Kannan, Gopal Radhakrishnan

2018-02-01

The research work on biodiesel becomes more attractive in the context of limited availability of petroleum fuels and rapid increase of harmful emissions from diesel engine using conventional fossil fuels. The present investigation has dealt with the influence of biodiesel-diesel-ethanol (diestrol) water micro emulsion fuel (B60D20E20M) on the performance, emission and combustion characteristics of a diesel engine under different injection pressure and timing. The results revealed that the maximum brake thermal efficiency of 32.4% was observed at an injection pressure of 260 bar and injection timing of 25.5°bTDC. In comparison with diesel, micro emulsion fuel showed reduction in carbon monoxide (CO) and total hydrocarbon (THC) by 40 and 24%, respectively. Further, micro emulsion fuel decreased nitric oxide (NO) emission and smoke emission by 7 and 20.7%, while the carbon dioxide (CO2) emission is similar to that of diesel.

Analysis of maximum allowable fragment heights during dissolution of high flux isotope reactor fuel in an h-canyon dissolver

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

Daniel, G.; Rudisill, T.

2017-07-17

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3O 8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2. The HFIR fuel cores will be dissolved using a flowsheet developed by the Savannahmore » River National Laboratory (SRNL) in either the 6.4D or 6.1D dissolver using a unique insert. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The recovered U will be down-blended into low-enriched U for subsequent use as commercial reactor fuel. During the development of the HFIR fuel dissolution flowsheet, the cycle time for the initial core was estimated at 28 to 40 h. Once the cycle is complete, H-Canyon personnel will open the dissolver and probe the HFIR insert wells to determine the height of any fuel fragments which did not dissolve. Before the next core can be charged to the dissolver, an analysis of the potential for H 2 gas generation must show that the combined surface area of the fuel fragments and the subsequent core will not generate H 2 concentrations in the dissolver offgas which exceeds 60% of the lower flammability limit (LFL) of H 2 at 200 °C. The objective of this study is to identify the maximum fuel fragment height as a function of the Al concentration in the dissolving solution which will provide criteria for charging successive HFIR cores to an H-Canyon dissolver.« less

Nuclear fuel pin scanner

DOEpatents

Bramblett, Richard L.; Preskitt, Charles A.

1987-03-03

Systems and methods for inspection of nuclear fuel pins to determine fiss loading and uniformity. The system includes infeed mechanisms which stockpile, identify and install nuclear fuel pins into an irradiator. The irradiator provides extended activation times using an approximately cylindrical arrangement of numerous fuel pins. The fuel pins can be arranged in a magazine which is rotated about a longitudinal axis of rotation. A source of activating radiation is positioned equidistant from the fuel pins along the longitudinal axis of rotation. The source of activating radiation is preferably oscillated along the axis to uniformly activate the fuel pins. A detector is provided downstream of the irradiator. The detector uses a plurality of detector elements arranged in an axial array. Each detector element inspects a segment of the fuel pin. The activated fuel pin being inspected in the detector is oscillated repeatedly over a distance equal to the spacing between adjacent detector elements, thereby multiplying the effective time available for detecting radiation emissions from the activated fuel pin.

2D and 3D Modeling Efforts in Fuel Film Cooling of Liquid Rocket Engines (Conference Paper with Briefing Charts)

DTIC Science & Technology

2017-01-12

J =0.13%, FFC Varies (b) J =0.13%, Main chamber Varies ( c ) J ...b) J =0.13%, Main Varies ( c ) J =1.23%, FFC Varies (d) J =1.18%, Main Varies (e) J =7.89%, FFC Varies (f) J =7.34%, Main Varies Figure 8. Power Spectral...Statement A: Approved for Public Release; Distribution Unlimited. PA Clearance #16569 (a) J =0.13%, FFC Varies (b) J =0.13%, Main Varies ( c ) J =2.79%,

Pyrolysis result of polyethylene waste as fuel for solid oxide fuel cell with samarium doped-ceria (SDC)-carbonate as electrolyte

NASA Astrophysics Data System (ADS)

Syahputra, R. J. E.; Rahmawati, F.; Prameswari, A. P.; Saktian, R.

2017-02-01

In this research, the result of pyrolysis on polyethylene was used as fuel for a solid oxide fuel cell (SOFC). The pyrolysis result is a liquid which consists of hydrocarbon chains. According to GC-MS analysis, the hydrocarbons mainly consist of C7 to C20 hydrocarbon chain. Then, the liquid was applied to a single cell of NSDC-L | NSDC | NSDC-L. NSDC is a composite SDC (samarium doped-ceria) with sodium carbonate. Meanwhile, NSDC-L is a composite of NSDC with LiNiCuO (LNC). NSDC and LNC were analyzed by X-ray diffraction to understand their crystal structure. The result shows that presence of carbonate did not change the crystal structure of SDC. SEM EDX analysis for fuel cell before and after being loaded with polyethylene oil to get information of element diffusion to the electrolyte. Meanwhile, the conductivity properties were investigated through impedance measurement. The presence of carbonate even increases the electrical conductivity. The single cell test with the pyrolysis result of polyethylene at 300 - 600 °C, found that the highest power density is at 600 °C with the maximum power density of 0.14 mW/cm2 and open circuit voltage of 0.4 Volt. Elemental analysis at three point spots of single cell NDSC-L |NSDC|NSDC-L found that a migration of ions was occurred during fuel operation at 300 - 600 °C.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

Effects of reduced nocturnal temperature on pig performance and energy consumption in swine nursery rooms.

PubMed

Johnston, L J; Brumm, M C; Moeller, S J; Pohl, S; Shannon, M C; Thaler, R C

2013-07-01

The objective of this investigation was to determine the effect of a reduced nocturnal temperature (RNT) regimen on performance of weaned pigs and energy consumption during the nursery phase of production. The age of weaned pigs assigned to experiments ranged from 16 to 22 d. In Exp. 1, 3 stations conducted 2 trials under a common protocol that provided data from 6 control rooms (CON; 820 pigs) and 6 RNT rooms (818 pigs). Two mirror-image nursery rooms were used at each station. Temperature in the CON room was set to 30°C for the first 7 d, then reduced by 2°C per week through the remainder of the experiment. Room temperature settings were held constant throughout the day and night. The temperature setting in the RNT room was the same as CON during the first 7 d, but beginning on the night of d 7, the room temperature setting was reduced 6°C from the daytime temperature from 1900 to 0700 h. The use of heating fuel and electricity were measured weekly in each room. Overall, ADG (0.43 kg), ADFI (0.62 kg), and G:F (0.69) were identical for CON and RNT rooms. Consumption of heating fuel [9,658 vs. 7,958 British thermal units (Btu)·pig(-1)·d(-1)] and electricity (0.138 vs. 0.125 kilowatt-hour (kWh)·pig(-1)·d(-1)] were not statistically different for CON and RNT rooms, respectively. In Exp. 2, 4 stations conducted at least 2 trials that provided data from 9 CON rooms (2,122 pigs) and 10 RNT rooms (2,176 pigs). Experimental treatments and protocols were the same as Exp. 1, except that the RNT regimen was imposed on the night of d 5 and the targeted nighttime temperature reduction was 8.3°C. Neither final pig BW (21.8 vs. 21.5 kg; SE = 0.64), ADG (0.45 vs. 0.44 kg; SE = 0.016), ADFI (0.61 vs. 0.60 kg; SE = 0.019), nor G:F (0.75 vs. 0.75; SE = 0.012) were different for pigs housed in CON or RNT rooms, respectively. Consumption of heating fuel and electricity was consistently reduced in RNT rooms for all 4 stations. Consumption of heating fuel (10,019 vs. 7,061 Btu·pig(-1)·d(-1); SE = 1,467) and electricity (0.026 vs. 0.021 kWh·pig-1·d-1; SE = 0.004) were lower (P < 0.05) in the RNT rooms compared with CON rooms. This represents a 30% reduction in heating fuel use and a 20% reduction in electrical use with no differences in pig growth performance or health. From these experiments, we conclude that imposing a RNT regimen from 1900 to 0700 h is effective in reducing energy costs in the nursery without compromising pig performance, which will reduce production costs and decrease emissions of greenhouse gases.

REACTOR FUEL ELEMENTS TESTING CONTAINER

DOEpatents

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Sensitivity Analysis for Multidisciplinary Systems (SAMS)

DTIC Science & Technology

2016-12-01

support both mode-based structural representations and time-dependent, nonlinear finite element structural dynamics. This interim report describes...Adaptation, & Sensitivity Toolkit • Elasticity, heat transfer, & compressible flow • Adjoint solver for sensitivity analysis • High-order finite elements ...PROGRAM ELEMENT NUMBER 62201F 6. AUTHOR(S) Richard D. Snyder 5d. PROJECT NUMBER 2401 5e. TASK NUMBER N/A 5f. WORK UNIT NUMBER Q1FS 7

Anaerobic biodegradation of biofuels and their impact on the corrosion of a Cu-Ni alloy in marine environments.

PubMed

Liang, Renxing; Aydin, Egemen; Le Borgne, Sylvie; Sunner, Jan; Duncan, Kathleen E; Suflita, Joseph M

2018-03-01

Fuel biodegradation linked to sulfate reduction can lead to corrosion of the metallic infrastructure in a variety of marine environments. However, the biological stability of emerging biofuels and their potential impact on copper-nickel alloys commonly used in marine systems has not been well documented. Two potential naval biofuels (Camelina-JP5 and Fisher-Tropsch-F76) and their petroleum-derived counterparts (JP5 and F76) were critically assessed in seawater/sediment incubations containing a metal coupon (70/30 Cu-Ni alloy). Relative to a fuel-unamended control (1.2 ± 0.4 μM/d), Camelina-JP5 (86.4 ± 1.6 μM/d) and JP5 (77.6 ± 8.3 μM/d) stimulated much higher rates of sulfate reduction than either FT-F76 (11.4 ± 2.7 μM/d) or F76 (38.4 ± 3.7 μM/d). The general corrosion rate (r 2  = 0.91) and pitting corrosion (r 2  = 0.92) correlated with sulfate loss in these incubations. Despite differences in microbial community structure on the metal or in the aqueous or sediment phases, sulfate reducing bacteria affiliated with Desulfarculaceae and Desulfobacteraceae became predominant upon fuel amendment. The identification of alkylsuccinates and alkylbenzylsuccinates attested to anaerobic metabolism of fuel hydrocarbons. Sequences related to Desulfobulbaceae were highly enriched (34.2-64.8%) on the Cu-Ni metal surface, regardless of whether the incubation received a fuel amendment. These results demonstrate that the anaerobic metabolism of biofuel linked to sulfate reduction can exacerbate the corrosion of Cu-Ni alloys. Given the relative lability of Camelina-JP5, particular precaution should be taken when incorporating this hydroprocessed biofuel into marine environments serviced by a Cu-Ni metallic infrastructure. Copyright © 2017 Elsevier Ltd. All rights reserved.

RER SPECTRA OBTAINED WITH A MULTICRYSTAL SPECTROMETER

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

Austin, W.E.; Champion, W.R.

1959-11-01

Relative gamma spectra were obtained twenty feet from the Hadiation Effects Reactor. The measurements were made using a multicry-stal spectrometer. This design incorporates pair and anticompton spectrometers in combination. Two reactor configurations were used; with shield tanks empty- and water filled. The spectra were obtained before the fuel elements were run at high power. Consequently very little of the fission product spectrum is tntermined. (J.R.D.)

Determination of trace elements in automotive fuels by filter furnace atomic absorption spectrometry

NASA Astrophysics Data System (ADS)

Anselmi, Anna; Tittarelli, Paolo; Katskov, Dmitri A.

2002-03-01

The determination of Cd, Cr, Cu, Pb and Ni was performed in gasoline and diesel fuel samples by electrothermal atomic absorption spectrometry using the Transverse Heated Filter Atomizer (THFA). Thermal conditions were experimentally defined for the investigated elements. The elements were analyzed without addition of chemical modifiers, using organometallic standards for the calibration. Forty-microliter samples were injected into the THFA. Gasoline samples were analyzed directly, while diesel fuel samples were diluted 1:4 with n-heptane. The following characteristic masses were obtained: 0.8 pg Cd, 6.4 pg Cr, 12 pg Cu, 17 pg Pb and 27 pg Ni. The limits of determination for gasoline samples were 0.13 μg/kg Cd, 0.4 μg/kg Cr, 0.9 μg/kg Cu, 1.5 μg/kg Pb and 2.5 μg/kg Ni. The corresponding limit of determination for diesel fuel samples was approximately four times higher for all elements. The element recovery was performed using the addition of organometallic compounds to gasoline and diesel fuel samples and was between 85 and 105% for all elements investigated.

Concentrations of Trace Elements in Settling Particulate Matter and Particulate Element Fluxes in the Concepción Bay, the Gulf of California during the years 1996-1998

NASA Astrophysics Data System (ADS)

Leal Acosta, M. L.; Choumiline, E.; Silverberg, N.

2009-12-01

María Luisa Leal Acosta, Evgueni Choumiline y Norman Silverberg National Polytechnic Institute, Interdisciplinary Center of Marine Sciences Av. Instituto Politécnico Nacional s/n Col. Palo de Santa Rita Apdo. Postal 592 23096 La Paz, B.C.S., México. The Concepcion Bay on the Peninsula de Baja California outstands by its high biological productivity and shallow gas-hydrothermal discharges. During a summer season the anoxic and hypoxic conditions frequently appear in the water column. To know the biogeochemistry of elements in this bay, a preliminary study of the composition of settling particulate matter (SPM) was carried out. The samples were collected since May 22, 1996 to January 30, 1997 (first period with low fluxes) and during January 31, 1997 - March 22, 1998 (second period with high fluxes) using a trap, anchored at south of the bay. The major element (Fe and Ca) and 21 trace element contents in SPM were determined by instrumental neutron activation analysis. The average lithogenic contribution in the particulate calculated using Sc is terrigenous indicator 39% and 19% for the first and second period respectively. The As, Br, Cr, Sb, Se, Sr y U average concentrations in the SPM were higher than their average crustal abundances. The principal component analysis shows the existents of three associations of the elements in settling particles: a) lithogenic (As, Co, Cr, Fe, lantanides and Sc); b) biogenic (Sr and Ca); c) authigenic (U and Se). The average element enrichment factors were higher in the first period of SPM sampling from: Se (739)> Zn (523)> Cr(105)> Br(104)> Sb(97)> As (69). The normalization of the rare earth elements contents in SPM with North American Shale Composite allowed to establish the negative europium anomaly in 1996 samples, and second period was characterized by positive europium anomaly, also detected for the sediments beneath the trap. Average fluxes of particulate elements during this period correspond to a following sequence: Fe (8.9 mg m-2 d-1)> Ca (6.9 mg m-2 d-1)> Sb (3.2 μg m-2 d-1)>Cr (2.1 μg m-2 d-1)> Sr (0.3 μg m-2 d-1)> Br (48.2 ng m-2 d-1)> U (27.9 ng m-2 d-1)> As (21.9 ng m-2 d-1) > Co (17.9 ng m-2 d-1)> Ce (17.1 ng m-2 d-1)> Nd (9.8 ng m-2 d-1)> La (7.6 ng m-2 d-1)> Se (7.4 ng m-2 d-1)> Rb (5.7 ng m-2 d-1)> Sc (4.4 ng m-2 d-1)> Sm (3.0 ng m-2 d-1)> Th (2.9 ng m-2 d-1)> Cs (1.8 ng m-2 d-1)> Hf (1.2 ng m-2 d-1)> Tb (0.6 ng m-2 d-1)> Lu (0.4 ng m-2 d-1)> Yb (0.4 ng m-2 d-1)> Eu (0.3 ng m-2 d-1).

Alternative Fuels for use in DoD/Army Tactical Ground Systems

DTIC Science & Technology

2011-02-03

Documentation Page Form ApprovedOMB No. 0704-0188 Public reporting burden for the collection of information is estimated to average 1 hour per response...for failing to comply with a collection of information if it does not display a currently valid OMB control number. 1 . REPORT DATE 03 FEB 2011 2...and approval of new fuels • What has been done so far – some examples – TRL 1 -4: Fuel properties – TRL 5-6: Component / engine evaluations – TRL 7-8

FUEL ELEMENT

DOEpatents

Bean, R.W.

1963-11-19

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

DART model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

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

Rest, J.; Hofman, G.L.

1997-12-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas-induced fuel swelling, interaction of fuel with the matrix aluminum, for the resultant reaction-product swelling, and for the calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data.

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.

MEANS FOR COOLING REACTORS

DOEpatents

Wheeler, J.A.

1957-11-01

A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

Affordable Development and Demonstration of a Small NTR Engine and Stage: How Small is Big Enough?

NASA Technical Reports Server (NTRS)

Borowski, S. K.; Sefcik, R. J.; Fittje, J. E.; McCurdy, D. R.; Qualls, A. L.; Schnitzler, B. G.; Werner, J.; Weitzberg, A.; Joyner, C. R.

2015-01-01

In FY11, NASA formulated a plan for Nuclear Thermal Propulsion (NTP) development that included Foundational Technology Development followed by system-level Technology Demonstrations The ongoing NTP project, funded by NASAs Advanced Exploration Systems (AES) program, is focused on Foundational Technology Development and includes 5 key task activities:(1) Fuel element fabrication and non-nuclear validation testing of heritage fuel options;(2) Engine conceptual design;(3) Mission analysis and engine requirements definition;(4) Identification of affordable options for ground testing; and(5) Formulation of an affordable and sustainable NTP development program Performance parameters for Point of Departure designs for a small criticality-limited and full size 25 klbf-class engine were developed during FYs 13-14 using heritage fuel element designs for both RoverNERVA Graphite Composite (GC) and Ceramic Metal (Cermet) fuel forms To focus the fuel development effort and maximize use of its resources, the AES program decided, in FY14, that a leader-follower down selection between GC and cermet fuel was required An Independent Review Panel (IRP) was convened by NASA and tasked with reviewing the available fuel data and making a recommendation to NASA. In February 2015, the IRP recommended and the AES program endorsed GC as the leader fuel In FY14, a preliminary development schedule DDTE plan was produced by GRC, DOE industry for the AES program. Assumptions, considerations and key task activities are presented here Two small (7.5 and 16.5 klbf) engine sizes were considered for ground and flight technology demonstration within a 10-year timeframe; their ability to support future human exploration missions was also examined and a recommendation on a preferred size is provided.

Analysis of Resonance Response Performance of C-Band Antenna Using Parasitic Element

PubMed Central

Islam, M. T.; Misran, N.; Mandeep, J. S.

2014-01-01

Analysis of the resonance response improvement of a planar C-band (4–8 GHz) antenna is proposed using parasitic element method. This parasitic element based method is validated for change in the active and parasitic antenna elements. A novel dual-band antenna for C-band application covering 5.7 GHz and 7.6 GHz is designed and fabricated. The antenna is composed of circular parasitic element with unequal microstrip lines at both sides and a rectangular partial ground plane. A fractional bandwidth of 13.5% has been achieved from 5.5 GHz to 6.3 GHz (WLAN band) for the lower band. The upper band covers from 7.1 GHz to 8 GHz with a fractional bandwidth of 12%. A gain of 6.4 dBi is achieved at the lower frequency and 4 dBi is achieved at the upper frequency. The VSWR of the antenna is less than 2 at the resonance frequency. PMID:24895643

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas L.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

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.

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

DOEpatents

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

JACKETED FUEL ELEMENT

DOEpatents

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

1959-02-01

These fuel elements are comprised of a homogeneous metallic uranium body completely enclosed and sealed in an aluminum cover. The uranium body and aluminum cover are bonded together by a layer of zinc located between them. The bonding layer serves to improve transfer of heat, provides an additional protection against corrosion of the uranium by the coolant, and also localizes any possible corrosion by preventing travel of corrosive material along the surface of the fuel element.

Comprehensive Fuel Spray Modeling and Impacts on Chamber Acoustics in Combustion Dynamics Simulations

DTIC Science & Technology

2013-05-01

multiple swirler configurations and fuel injector locations at atmospheric pressure con- ditions. Both single-element and multiple-element LDI...the swirl number, Reynolds’ number and injector location in the LDI element. Besides the multi-phase flow characteristics, several experimen- tal...region downstream of the fuel injector on account of a sta- ble and compact precessing vortex core. Recent ex- periments conducted by the Purdue group have

FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

DOEpatents

Flint, O.

1961-01-10

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

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

Lethal and behavioral impacts of diesel and fuel oil on the Antarctic amphipod Paramoera walkeri.

PubMed

Brown, Kathryn E; King, Catherine K; Harrison, Peter L

2017-09-01

Toxicity testing with Antarctic species is required for risk assessment of fuel spills in Antarctic coastal waters. The lethal and sublethal (movement behavior) sensitivities of adults and juveniles of the Antarctic amphipod Paramoera walkeri to the water accommodated fractions (WAFs) of 3 fuels were estimated in extended-duration tests at -1 °C to 21 d. Response of P. walkeri for lethal hydrocarbon concentrations was slow, with 50% lethal concentrations (LC50s) first able to be estimated at 7 d for adults exposed to Special Antarctic Blend diesel (SAB), which had the highest hydrocarbon concentrations of the 3 fuel WAFs. Juveniles showed greater response to marine gas oil (MGO) and intermediate residual fuel oil (IFO 180) at longer exposure durations and were most sensitive at 21 d to IFO 180 (LC50 = 12 μg/L). Adults were initially more sensitive than juveniles; at 21 d, however, juveniles were more than twice as sensitive as adults to SAB (LC50 = 153 μg/L and 377 μg/L, respectively). Significant effects on movement behavior were evident at earlier time points and lower concentrations than was mortality in all 3 fuel WAFs, and juveniles were highly sensitive to sublethal effects of MGO. These first estimates of Antarctic amphipod sensitivity to diesel and fuel oils in seawater contribute to the development of ecologically relevant risk assessments for management of hydrocarbon contamination in the region. Environ Toxicol Chem 2017;36:2444-2455. © 2017 SETAC. © 2017 SETAC.

Biodegradation of biodiesel fuels

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

Zhang, X.; Haws, R.; Wright, B.

1995-12-31

Biodiesel fuel test substances Rape Ethyl Ester (REE), Rape Methyl Ester (RME), Neat Rape Oil (NR), Say Methyl Ester (SME), Soy Ethyl Ester (SEE), Neat Soy Oil (NS), and proportionate combinations of RME/diesel and REE/diesel were studied to test the biodegradability of the test substances in an aerobic aquatic environment using the EPA 560/6-82-003 Shake Flask Test Method. A concurrent analysis of Phillips D-2 Reference Diesel was also performed for comparison with a conventional fuel. The highest rates of percent CO{sub 2} evolution were seen in the esterified fuels, although no significant difference was noted between them. Ranges of percentmore » CO{sub 2} evolution for esterified fuels were from 77% to 91%. The neat rape and neat soy oils exhibited 70% to 78% CO{sub 2} evolution. These rates were all significantly higher than those of the Phillips D-2 reference fuel which evolved from 7% to 26% of the organic carbon to CO{sub 2}. The test substances were examined for BOD{sub 5} and COD values as a relative measure of biodegradability. Water Accommodated Fraction (WAF) was experimentally derived and BOD{sub 5} and COD analyses were carried out with a diluted concentration at or below the WAF. The results of analysis at WAF were then converted to pure substance values. The pure substance BOD{sub 5} and COD values for test substances were then compared to a control substance, Phillips D-2 Reference fuel. No significant difference was noted for COD values between test substances and the control fuel. (p > 0.20). The D-2 control substance was significantly lower than all test substances for BCD, values at p << 0.01. RME was also significantly lower than REE (p < 0.05) and MS (p < 0.01) for BOD{sub 5} value.« less

Characterisation of a novel integrative and conjugative element ICESsD9 carrying erm(B) and tet(O) resistance determinants in Streptococcus suis, and the distribution of ICESsD9-like elements in clinical isolates.

PubMed

Huang, Kaisong; Song, Yajing; Zhang, Qiang; Zhang, Anding; Jin, Meilin

2016-12-01

This study identified a novel integrative and conjugative element (ICESsD9) carrying erm(B) and tet(O) resistance determinants in Streptococcus suis D9 and determined its prevalence in clinical isolates. Comparative genome analysis was performed using Mauve and Artemis Comparison Tool visualisation programs. Inverse PCR was utilised to detect its circular intermediate. The transfer capacity of ICESsD9 was evaluated by mating assays using S. suis A7 and Enterococcus faecalis JH2-2 as recipients. A genome walking approach was employed to analyse the characteristics of integration sites in transconjugants. A total of 118 clinical S. suis isolates were tested by PCR mapping assays to detect ICESsD9-like elements. MLST was performed on isolates containing ICESsD9 variants to determine their clonal relatedness. This 55 683-bp element can actively excise from the chromosome. Additionally, it was capable of transferring both into S. suis and E. faecalis with frequencies of 1.2×10 -4 and 5.8×10 -6 per donor, respectively. When investigating integration site features, it was found that ICESsD9 can enter S. suis and E. faecalis chromosomes by different sites, generating 15-bp and 3-bp direct repeat sequences, respectively. Twelve isolates mainly belonging to sequence types ST1, ST7 and ST28 were confirmed to harbour ICESsD9-like elements. In conclusion, this study provides the first description of an ICE in S. suis that is capable of transferring both into S. suis and E. faecalis. The presence of different ICESsD9 variants in clinical isolates suggests already wide dissemination of this family element in S. suis in China. Copyright © 2016 International Society for Chemotherapy of Infection and Cancer. Published by Elsevier Ltd. All rights reserved.

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

Sarta, Jose A.; Castiblanco, Luis A

With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer ofmore » the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.« less

Experimental and Numerical Investigation of Pressure Drop in Silicon Carbide Fuel Rod for Application in Pressurized Water Reactors

NASA Astrophysics Data System (ADS)

Abir, Ahmed Musafi

Spacer grids are used in Pressurized Water Reactors (PWRs) fuel assemblies which enhances heat transfer from fuel rods. However, there remain regions of low turbulence in between the spacer grids. To enhance turbulence in these regions surface roughness is applied on the fuel rod walls. Meyer [1] used empirical correlations to predict heat transfer and friction factor for artificially roughened fuel rod bundles at High Performance Light Water Reactors (LWRs). Their applicability was tested by Carrilho at University of South Carolina's (USC) Single Heated Element Loop Tester (SHELT). He attained a heat transfer and friction factor enhancement of 50% and 45% respectively, using Inconel nuclear fuel rods with square transverse ribbed surface. Following him Najeeb conducted a similar study due to three dimensional diamond shaped blocks in turbulent flow. He recorded a maximum heat transfer enhancement of 83%. At present, several types of materials are being used for fuel rod cladding including Zircaloy, Uranium oxide, etc. But researchers are actively searching for new material that can be a more practical alternative. Silicon Carbide (SiC) has been identified as a material of interest for application as fuel rod cladding [2]. The current study deals with the experimental investigation to find out the friction factor increase of a SiC fuel rod with 3D surface roughness. The SiC rod was tested at USC's SHELT loop. The experiment was conducted in turbulent flowing Deionized (DI) water at steady state conditions. Measurements of Flow rate and pressure drop were made. The experimental results were also validated by Computational Fluid Dynamics (CFD) analysis in ANSYS Fluent. To simplify the CFD analysis and to save computational resources the 3D roughness was approximated as a 2D one. The friction factor results of the CFD investigation was found to lie within +/-8% of the experimental results. A CFD model was also run with the energy equation turned on, and a heat generation of 8 kW applied to the rod. A maximum heat transfer enhancement of 18.4% was achieved at the highest flow rate investigated (i.e. Re=109204).

A finite element analysis modeling tool for solid oxide fuel cell development: coupled electrochemistry, thermal and flow analysis in MARC®

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

Khaleel, Mohammad A.; Lin, Zijing; Singh, Prabhakar

2004-05-03

A 3D simulation tool for modeling solid oxide fuel cells is described. The tool combines the versatility and efficiency of a commercial finite element analysis code, MARC{reg_sign}, with an in-house developed robust and flexible electrochemical (EC) module. Based upon characteristic parameters obtained experimentally and assigned by the user, the EC module calculates the current density distribution, heat generation, and fuel and oxidant species concentration, taking the temperature profile provided by MARC{reg_sign} and operating conditions such as the fuel and oxidant flow rate and the total stack output voltage or current as the input. MARC{reg_sign} performs flow and thermal analyses basedmore » on the initial and boundary thermal and flow conditions and the heat generation calculated by the EC module. The main coupling between MARC{reg_sign} and EC is for MARC{reg_sign} to supply the temperature field to EC and for EC to give the heat generation profile to MARC{reg_sign}. The loosely coupled, iterative scheme is advantageous in terms of memory requirement, numerical stability and computational efficiency. The coupling is iterated to self-consistency for a steady-state solution. Sample results for steady states as well as the startup process for stacks with different flow designs are presented to illustrate the modeling capability and numerical performance characteristic of the simulation tool.« less

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

Potter, David Charles; Taylor, Craig Michael; Coons, James Elmer

The percent void of the Fort Saint Vrain (FSV) material is estimated to be 21.1% based on the volume of the gap at the top of the drums, the volume of the coolant channels in the FSV fuel element, and the volume of the fuel handling channel in the FSV fuel element.

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.

Advances in Materials and System Technology for Portable Fuel Cells

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R.

2007-01-01

This viewgraph presentation describes the materials and systems engineering used for portable fuel cells. The contents include: 1) Portable Power; 2) Technology Solution; 3) Portable Hydrogen Systems; 4) Direct Methanol Fuel Cell; 5) Direct Methanol Fuel Cell System Concept; 6) Overview of DMFC R&D at JPL; 7) 300-Watt Portable Fuel Cell for Army Applications; 8) DMFC units from Smart Fuel Cell Inc, Germany; 9) DMFC Status and Prospects; 10) Challenges; 11) Rapid Screening of Well-Controlled Catalyst Compositions; 12) Screening of Ni-Zr-Pt-Ru alloys; 13) Issues with New Membranes; 14) Membranes With Reduced Methanol Crossover; 15) Stacks; 16) Hybrid DMFC System; 17) Small Compact Systems; 18) Durability; and 19) Stack and System Parameters for Various Applications.

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-01-14

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-05-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

DOEpatents

Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

1963-09-01

A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

Effect of secondary fuels and combustor temperature on mercury speciation in pulverized fuel co-combustion: part 1

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

Shishir P. Sable; Wiebren de Jong; Ruud Meij

2007-08-15

The present work mainly involves bench scale studies to investigate partitioning of mercury in pulverized fuel co-combustion at 1000 and 1300{sup o}C. High volatile bituminous coal is used as a reference case and chicken manure, olive residue, and B quality (demolition) wood are used as secondary fuels with 10 and 20% thermal shares. The combustion experiments are carried out in an entrained flow reactor with a fuel input of 7-8 kWth. Elemental and total gaseous mercury concentrations in the flue gas of the reactor are measured on-line, and ash is analyzed for particulate mercury along with other elemental and surfacemore » properties. Animal waste like chicken manure behaves very differently from plant waste. The higher chlorine contents of chicken manure cause higher ionic mercury concentrations whereas even with high unburnt carbon, particulate mercury reduces with increase in the chicken manure share. This might be a problem due to coarse fuel particles, low surface area, and iron contents. B-wood and olive residue cofiring reduces the emission of total gaseous mercury and increases particulate mercury capture due to unburnt carbon formed, fine particles, and iron contents of the ash. Calcium in chicken manure does not show any effect on particulate or gaseous mercury. It is probably due to a higher calcium sulfation rate in the presence of high sulfur and chlorine contents. However, in plant waste cofiring, calcium may have reacted with chlorine to reduce ionic mercury to its elemental form. According to thermodynamic predictions, almost 50% of the total ash is melted to form slag at 1300{sup o}C in cofiring because of high calcium, iron, and potassium and hence mercury and other remaining metals are concentrated in small amounts of ash and show an increase at higher temperatures. No slag formation was predicted at 1000{sup o}C. 24 refs., 8 figs., 4 tabs.« less

Fuel shipment experience, fuel movements from the BMI-1 transport cask

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

Bauer, Thomas L.; Krause, Michael G

1986-07-01

The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including severalmore » factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask.« less

The Finite Element Modelling and Dynamic Characteristics Analysis about One Kind of Armoured Vehicles’ Fuel Tanks

NASA Astrophysics Data System (ADS)

Gao, Yang; Ge, Zhishang; Zhai, Weihao; Tan, Shiwang; Zhang, Feng

2018-01-01

The static and dynamic characteristics of fuel tank are studied for the armoured vehicle in this paper. The CATIA software is applied to build the CAD model of the armoured vehicles’ fuel tank, and the finite element model is established in ANSYS Workbench. The finite element method is carried out to analyze the static and dynamic mechanical properties of the fuel tank, and the first six orders of mode shapes and their frequencies are also computed and given in the paper, then the stress distribution diagram and the high stress areas are obtained. The results of the research provide some references to the fuel tanks’ design improvement, and give some guidance for the installation of the fuel tanks on armoured vehicles, and help to improve the properties and the service life of this kind of armoured vehicles’ fuel tanks.

Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel

NASA Astrophysics Data System (ADS)

Lewis, operating defective fuel B. J.; Thompson, W. T.; Akbari, F.; Thompson, D. M.; Thurgood, C.; Higgs, J.

2004-07-01

A theoretical treatment has been developed to predict the fuel oxidation behaviour in operating defective nuclear fuel elements. The equilibrium stoichiometry deviation in the hyper-stoichiometric fuel has been derived from thermodynamic considerations using a self-consistent set of thermodynamic properties for the U-O system, which emphasizes replication of solubilities and three-phase invariant conditions displayed in the U-O binary phase diagram. The kinetics model accounts for multi-phase transport including interstitial oxygen diffusion in the solid and gas-phase transport of hydrogen and steam in the fuel cracks. The fuel oxidation model is further coupled to a heat conduction model to account for the feedback effect of a reduced thermal conductivity in the hyper-stoichiometric fuel. A numerical solution has been developed using a finite-element technique with the FEMLAB software package. The model has been compared to available data from several in-reactor X-2 loop experiments with defective fuel conducted at the Chalk River Laboratories. The model has also been benchmarked against an O/U profile measurement for a spent defective fuel element discharged from a commercial reactor.

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

DOEpatents

Moore, W.T.

1958-09-01

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

REACTOR UNLOADING

DOEpatents

Leverett, M.C.

1958-02-18

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

DOEpatents

Plott, R.F.

1958-10-28

A method of casting a protective jacket about a ura nium fuel element that will bond completely to the uranium without the use of stringers or supports that would ordinarily produce gaps in the cast metal coating and bond is presented. Preformed endcaps of alumlnum alloyed with 13% silicon are placed on the ends of the uranium fuel element. These caps will support the fuel element when placed in a mold. The mold is kept at a ing alloy but below that of uranium so the cast metal jacket will fuse with the endcaps forming a complete covering and bond to the fuel element, which would otherwise oxidize at the gaps or discontinuities lefi in the coating by previous casting methods.

URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

DOEpatents

Handwerk, J.H.; Noland, R.A.; Walker, D.E.

1957-09-10

In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

NUCLEAR REACTOR

DOEpatents

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

JACKETED FISSIONABLE MEMBER

DOEpatents

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

1960-09-13

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

U-Mo Plate Blister Anneal Interim Report

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

Francine J. Rice; Daniel M. Wachs; Adam B. Robinson

2010-10-01

Blister thresholds in fuel elements have been a longstanding performance parameter for fuel elements of all types. This behavior has yet to be fully defined for the RERTR U-Mo fuel types. Blister anneal studies that began in 2007 have been expanded to include plates from more recent RERTR experiments. Preliminary data presented in this report encompasses the early generations of the U-Mo fuel systems and the most recent but still developing fuel system. Included is an overview of relevant dispersion fuel systems for the purposes of comparison.

Production of a refined biooil derived by fast pyrolysis of chicken manure with chemical and physical characteristics close to those of fossil fuels.

PubMed

Monreal, Carlos M; Schnitzer, Morris

2011-01-01

The chemical and physical properties of raw biooils prevent their direct use in combustion engines. We processed raw pyrolytic biooil derived from chicken manure to yield a colorless refined biooil with diesel qualities. Chemical characterization of the refined biooil involved elemental and several spectroscopic analyses. The physical measurements employed were viscosity, density and heat of combustion. The elemental composition (% wt/wt) of the refined biooil was 82.7 % C, 15.3 % H, 0.2 % N and 1.8 % O, no S. Its viscosity was 0.006 Pa.s and a heat of combustion of 43 MJ kg(-1). The refined biooil fraction contains n-alkanes, ranging from n-C(14) to n-C(27), alkenes varying from C(10:1) to C(22:1), and long-chain alcohols. The refined biooil makes a good diesel fuel due to its chemical and physical properties.

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

Sublette, K.L.

With the continual increase in the utilization of high sulfur and high nitrogen containing fossil fuels, the release of airborne pollutants into the environment has become a critical problem. The fuel sulfur is converted to SO{sub 2} during combustion. Fuel nitrogen and a fraction of the nitrogen from the combustion air are converted to nitric oxide and nitrogen dioxide, NO{sub x}. For the past five years Combustion Engineering (now Asea Brown Boveri or ABB) and, since 1986, the University of Tulsa (TU) have been investigating the oxidation of H{sub 2}S by the facultatively anaerobic and autotrophic bacterium Thiobacillus denitrificans andmore » have developed a process, concept for the microbial removal of H{sub 2}S from a gas stream the simultaneous removal of SO{sub 2} and NO by D. desulfuricans and T. denitrificans co-cultures and cultures-in-series was demonstrated. These systems could not be sustained due to NO inhibition of D. desulfuricans. However, a preliminary economic analysis has shown that microbial reduction of SO{sub 2} to H{sub 2}S with subsequent conversion to elemental sulfur by the Claus process is both technically and economically feasible if a less expensive carbon and/or energy source can be found. It has also been demonstrated that T. denitrificans can be grown anaerobically on NO(g) as a terminal electron acceptor with reduction to elemental nitrogen. Microbial reduction of NO{sub x} is a viable process concept for the disposal of concentrated streams of NO{sub x} as may be produced by certain regenerable processes for the removal of SO{sub 2} and NO{sub x} from flue gas.« less

Intermetallic compounds of the heaviest elements and their homologs: the electronic structure and bonding of MM', where M=Ge, Sn, Pb, and element 114, and M'=Ni, Pd, Pt, Cu, Ag, Au, Sn, Pb, and element 114.

PubMed

Pershina, V; Anton, J; Fricke, B

2007-10-07

Fully relativistic (four-component) density-functional theory calculations were performed for intermetallic dimers MM', where M=Ge, Sn, Pb, and element 114, and MM'=group 10 elements (Ni, Pd, and Pt) and group 11 elements (Cu, Ag, and Au). PbM and 114M, where M are group 14 elements, were also considered. The results have shown that trends in spectroscopic properties-atomization energies D(e), vibrational frequencies omega(e), and bond lengths R(e), as a function of MM', are similar for compounds of Ge, Sn, Pb, and element 114, except for D(e) of PbNi and 114Ni. They were shown to be determined by trends in the energies and space distribution of the valence ns(MM')atomic orbitals (AOs). According to the results, element 114 should form the weakest bonding with Ni and Ag, while the strongest with Pt due to the largest involvement of the 5d(Pt) AOs. In turn, trends in the spectroscopic properties of MM' as a function of M were shown to be determined by the behavior of the np(1/2)(M) AOs. Overall, D(e) of the element 114 dimers are about 1 eV smaller and R(e) are about 0.2 a.u. larger than those of the corresponding Pb compounds. Such a decrease in bonding of the element 114 dimers is caused by the large SO splitting of the 7p orbitals and a decreasing contribution of the relativistically stabilized 7p(1/2)(114) AO. On the basis of the calculated D(e) for the dimers, adsorption enthalpies of element 114 on the corresponding metal surfaces were estimated: They were shown to be about 100-150 kJ/mol smaller than those of Pb.

A small, 1400 deg Kelvin, reactor for Brayton space power systems

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

The Shock and Vibration Digest, Volume 14, Number 4

DTIC Science & Technology

1982-04-01

temperature, humidity, shock, and vibration -- can influence this capability; as a result an almost continuous program of research and development has...pro- ducing reliability tests. For some time there has been interest in the Army Test Methodology program for developing a vibration system capable...geology of the Livermore Valley is obtained. 82-768 Transient Stress Wave Propagation in HTGR Fuel Element Impacts I.T. Almajan and P.D. Smith

French Flight Test Program LEA Status

DTIC Science & Technology

2010-09-01

RTO-EN-AVT-185 17 - 1 French Flight Test Program LEA Status Francois FALEMPIN MBDA France 1 avenue Reaumur Le Plessis Robinson FRANCE ...TITLE AND SUBTITLE French Flight Test Program LEA Status 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) 5d. PROJECT ...Bouchez, Nicolas Gascoin, Measurement for fuel reforming for scramjet thermal management: status of COMPARER project - AIAA-2009-7373. French

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

Stimpson, Shane G.

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA Core Simulator (VERA-CS) development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code being developed by the Idaho National Laboratory (INL) is the focus of ongoing work inmore » the Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework, BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, for this work, it uses a 2D azimuthally symmetric (R-Z) smeared-pellet model.« less

Thermochemical hydrogen production based on magnetic fusion

NASA Astrophysics Data System (ADS)

Krikorian, O. H.; Brown, L. C.

Preliminary results of a DoE study to define the configuration and production costs for a Tandem Mirror Reactor (TMR) heat source H2 fuel production plant are presented. The TMR uses the D-T reaction to produce thermal energy and dc electrical current, with an Li blanket employed to breed more H-3 for fuel. Various blanket designs are being considered, and the coupling of two of them, a heat pipe blanket to a Joule-boosted decomposer, and a two-temperature zone blanket to a fluidized bed decomposer, are discussed. The thermal energy would be used in an H2SO4 thermochemical cycler to produce the H2. The Joule-boosted decomposer, involving the use of electrically heated commercial SiC furnace elements to transfer process heat to the thermochemical H2 cycle, is found to yield H2 fuel at a cost of $12-14/GJ, which is the projected cost of fossil fuels in 30-40 yr, when the TMR H2 production facility would be operable.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Petrucco, Louis Jacob (Inventor); Daum, Wolfgang (Inventor)

2005-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Brown, Dale Marius (Inventor); Petrucco, Louis Jacob (Inventor); Lovett, Jeffery Allan (Inventor); Daum, Wolfgang (Inventor); Dunki-Jacobs, Robert John (Inventor)

2003-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Brown, Dale Marius (Inventor); Petrucco, Louis Jacob (Inventor); Lovett, Jeffery Allan (Inventor); Daum, Wolfgang (Inventor); Dunki-Jacobs, Robert John (Inventor)

1999-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Experimental investigation on the availability, performance, combustion and emission distinctiveness of bael oil/ diesel/ diethyl ether blends powered in a variable compression ratio diesel engine

NASA Astrophysics Data System (ADS)

Krishnamoorthi, M.; Malayalamurthi, R.

2018-02-01

The present work aims at experimental investigation on the combined effect of injection timing (IT) and injection pressure (IP) on the performance and emissions characteristics, and exergy analysis of a compression-ignition (CI) engine powered with bael oil blends. The tests were conducted using ternary blends of bael oil, diethyl ether (DEE) and neat diesel (D) at various engine loads at a constant engine speed (1500 rpm). With B2 (60%D + 30%bael oil+10%DEE) fuel, the brake thermal efficiency (BTE) of the engine is augmented by 3.5%, reduction of 4.7% of oxides of nitrogen (NOx) emission has been observed at 100% engine load with 250 bar IP. B2 fuel exhibits 7% lower scale of HC emissions compared to that of diesel fuel at 100% engine load in 23 °bTDC IT. The increment in both cooling water and exhaust gas availabilities lead to increasing exergy efficiency with increasing load. The exergy efficiency of about 62.17% has been recorded by B2 fuel at an injection pressure of 230 IP bar with 100% load. On the whole, B2 fuel displays the best performance and combustion characteristics. It also exhibits better characteristics of emissions level in terms of lower HC, smoke opacity and NOx.

Local Burn-Up Effects in the NBSR Fuel Element

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

Brown N. R.; Hanson A.; Diamond, D.

2013-01-31

This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peakingmore » relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.« less

A Search for New Fuel Components in Explosive Mixtures with Ammonium Nitrate

DTIC Science & Technology

1981-04-30

i AP-PENIMY A. rol~our Tnd#Ix entrl-s for dyes tested for euatecticmixture form7ation with wnmoniua.i nitrate 12010 C.!. Snivent tR.. d 3...lirainhto. %k ith sodium hydroide itnd sodium ýhlnrate l1r6nner, BP 739 182 o d )r nitrate (GP I1S52q) lh-lftier, Gi’ 3628P 9 (1’. 1, 308) (it) Iaat...K -__ __ __ __ _ R___ __ _ __ LEVEL... 0 Final Report A SEARCH FOR NEW FUEL COMPONENTS IN EXPLOSIVE MIXTURES WITH AMMONIUM NITRATE -m i Dr. Maurice C

Renewable Energy Production from DoD Installation Solid Wastes By Anaerobic Digestion

DTIC Science & Technology

2016-08-06

favorable environmental conditions including a mesophilic (37 oC) or thermophilic (55 oC) temperature , the absence of oxygen , and a pH between 6.5...a high temperature process that uses oxygen -starved combustion to convert dry organic matter to a syngas. Syngas is a low BTU fuel that can be used...production rates are at the 36.7 °C digester temperature . Parameter Units 7gCOD/L-d 12gCOD/L-d Effective SRT days 18.5 10.8 COD Conversion Efficiency % 67

Renewable Energy Production from DoD Installation Solid Wastes by Anaerobic Digestion

DTIC Science & Technology

2016-06-01

favorable environmental conditions including a mesophilic (37 oC) or thermophilic (55 oC) temperature , the absence of oxygen , and a pH between 6.5...a high temperature process that uses oxygen -starved combustion to convert dry organic matter to a syngas. Syngas is a low BTU fuel that can be used...production rates are at the 36.7 °C digester temperature . Parameter Units 7gCOD/L-d 12gCOD/L-d Effective SRT days 18.5 10.8 COD Conversion Efficiency % 67

Data summary report for fission product release Test VI-7

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

Osborne, M.F.; Lorentz, R.A.; Travis, J.R.

Test VI-7 was the final test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the Monticello boiling water reactor (BWR). The fuel had experienced a burnup of {approximately}-40 Mwd/kg U. It was heated in an induction furnace for successive 20-min periods at 2000 and 2300 K in a moist air-helium atmosphere. Integral releases were 69% for {sup 85}Kr, 52% for {sup 125}Sb, 71% for both {sup 134}Cs and {sup 137}Cs, and 0.04% for {sup 154}Eu. For the non-gamma-emitting species, release values for 42% for I, 4.1% formore » Ba, 5.3% for Mo, and 1.2% for Sr were determined. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.89 g, with 37% being collected on the thermal gradient tubes and 63% downstream on filters. Posttest examination of the fuel specimen indicated that most of the cladding was completely oxidized to ZrO{sub 2}, but that oxidation was not quite complete at the upper end. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL-Booth Model.« less

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL

DOEpatents

Draley, J.E.; Ruther, W.E.

1959-01-27

Fuel elements and coolant tubes used in nuclear reactors of the heterogeneous, water-cooled type are described, wherein the coolant tubes extend through the moderator and are adapted to contain the fuel elements. The invention comprises forming the coolant tubes and the fuel element cladding material from an alloy of aluminum and nickel, or an alloy of aluminum, nickel, alloys are selected to prevent intergranular corrosion of these components by water at temperatures up to 35O deg C.

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.

A REACTOR DESIGN PARAMETER STUDY

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

Fox, A.H.; LaVerne, M.E.; Burtnette, C.S.

1954-06-25

Multigroup calculations were performed on reflectormoderated systems to establish some of the nuclear characteristics of various reflector geometries and materials. C, Li/sup 7/, Li/sup 7/OD, and NaOD moderators were used with NaF-UF/ sub 4/ fuel. The results are tabulated for 57 moderator and dimensional variations. (D.E.B.)

FUEL ELEMENT

DOEpatents

Fortescue, P.; Zumwalt, L.R.

1961-11-28

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

40 CFR 86.113-04 - Fuel specifications.

Code of Federal Regulations, 2013 CFR

2013-07-01

... (CONTINUED) CONTROL OF EMISSIONS FROM NEW AND IN-USE HIGHWAY VEHICLES AND ENGINES Emission Regulations for..., Research, Min. D 2699 93 Sensitivity, Min. 7.5 Lead (organic), max. g/U.S. gal. (g/liter) D 3237 0.050 (0... sensitivity is defined as the Research octane number minus the Motor octane number. (iii) The Reid Vapor...

40 CFR 86.113-04 - Fuel specifications.

Code of Federal Regulations, 2011 CFR

2011-07-01

... (CONTINUED) CONTROL OF EMISSIONS FROM NEW AND IN-USE HIGHWAY VEHICLES AND ENGINES Emission Regulations for..., Research, Min. D 2699 93 Sensitivity, Min. 7.5 Lead (organic), max. g/U.S. gal. (g/liter) D 3237 0.050 (0... sensitivity is defined as the Research octane number minus the Motor octane number. (iii) The Reid Vapor...

FUEL ELEMENT CONSTRUCTION

DOEpatents

Simnad, M.T.

1961-08-15

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

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

2011-01-01

Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR JULY 1, 1959- SEPTEMBER 30, 1959

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

None

1960-10-31

ABS>Fuel Element Research and Development. Dynamic and static corrosion tests on 8001 Al were completed. Annealmmmg of 1100 cladding on 5083 and M400 cladding on X2219 were tested at 500 deg C, and investigation continued on producing X8101 Al alloy cladding in tube plates by extrusion. Boiler fuel element capsule irradiation tests and subassembly tests are described Heat transfer loop studies and fuel fabrication for the critical facility are reported. Boiler fuel element mechanical design and testing progress is desc ribed. and the superheater fuel element temperature evaluating routine is discussed. Low- enrichment superheater fuel element development included design studiesmore » and stainless steel powder and UO/sub 2/ powder fabrication studies Reactor Mechanical Studies. Research is reported on vessel and structure design, fabrication, and testing, recirculation system design, steam separator tests, and control rod studies. Nuclear Analysis. Reactor physics studies are reported on nuclear constants, baffle plate analysis, comparison of core representations, delayed neutron fraction. and shielding analysis of the reactor building. Reactor and system dynamics and critical experiments were also studied. Chemistry. Progress is reported on recombiner. radioactive gas removal and storage, ion exchanger and radiochemical processing. (For preceding period see ACNP-5915.) (T.R.H.)« less

Compact Fuel Element Environment Test

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

2012-01-01

Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

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

D. D. Keiser; J. I. Cole

2007-09-01

Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

DoD use of Domestically-Produced Alternative Fuels and Alternative Fuel Vehicles

DTIC Science & Technology

2014-04-10

based fuels and biodiesel , in DoD vehicles; (2) current and projected actions by the DoD to increase the use of alternative fuels in vehicles; (3) a...fuels and vehicles. 15. SUBJECT TERMS alternative fuel infrastructure, electric vehicles, biofuels, ethanol, biodiesel , drop-in, synthetic fuel...of: (1) use and potential use of domestically-produced alternative fuels including but not limited to, natural gas based fuels and biodiesel , in DoD

Preliminary neutronic analysis of a cavity test reactor

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

A reference configuration was calculated for a cavity test reactor to be used for testing the gascore nuclear rocket concept. A thermal flux of 4.1 x 10 to the 14th power neutrons per square centimeter per second in the cavity was provided by a driver fuel loading of 6.4 kg of enriched uranium in MTR fuel elements. The reactor was moderated and cooled by heavy water and reflected with 25.4 cm of beryllium. Power generation of 41.3 MW in the driver fuel is rejected to a heat sink. Design effort was directed toward minimization of driver power while maintaining 2.7 MW in the cavity during a test run. Ancillary data on material reactivity worths, reactivity coefficients, flux spectra, and power distributions are reported.

Dart model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

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

Rest, J.; Hofman, G.L.

1997-06-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas induced fuel swelling, interaction of fuel with the matrix aluminum, resultant reaction-product swelling, and calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data. DART results are compared with data for fuel swelling Of U{sub 3}SiAl-Al in plate, tube, and rod configurations as a function of fission density.more » Plate and tube calculations were performed at a constant fuel temperature of 373 K and 518 K, respectively. An irradiation temperature of 518 K results in a calculated aluminide layer thickness for the Russian tube that is in the center of the measured range (16 {mu}m). Rod calculations were performed with a temperature gradient across the rod characterized by surface and central temperatures of 373 K and 423 K, respectively. The effective yield stress of irradiated Al matrix material and the aluminide was determined by comparing the results of DART calculations with postirradiation immersion volume measurement of U{sub 3}SiAl plates. The values for the effective yield stress were used in all subsequent simulations. The lower calculated fuel swelling in the rod-type element is due to an assumed biaxial stress state. Fuel swelling in plates results in plate thickness increase only. Likewise, in tubes, only the wall thickness increases. Irradiation experiments have shown that plate-type dispersion fuel elements can develop blisters or pillows at high U-235 burnup when fuel compounds exhibiting breakaway swelling are used at moderate to high fuel volume fractions. DART-calculated interaction layer thickness and fuel swelling follows the trends of the observations. 3 refs., 2 figs.« less

Cosmochemistry

NASA Astrophysics Data System (ADS)

Esteban, C.; García López, R. J.; Herrero, A.; Sánchez, F.

2004-03-01

1. Primordial alchemy: from the Big Bang to the present Universe G. Steigman; 2. Stellar nucleosynthesis N. Langer; 3. Obervational aspects of stellar nucleosynthesis D. L. Lambert; 4. Abundance determinations in HII regions and planetary nebulae G. Stasinska; 5. Element abundances in nearby galaxies D. R. Garnett; 6. Chemical evolution of galaxies and intracluster medium F.Matteucci; 7. Element abundances through the cosmic ages M. Pettini.

Cosmochemistry

NASA Astrophysics Data System (ADS)

Esteban, C.; García López, R. J.; Herrero, A.; Sánchez, F.

2011-01-01

1. Primordial alchemy: from the Big Bang to the present Universe G. Steigman; 2. Stellar nucleosynthesis N. Langer; 3. Obervational aspects of stellar nucleosynthesis D. L. Lambert; 4. Abundance determinations in HII regions and planetary nebulae G. Stasinska; 5. Element abundances in nearby galaxies D. R. Garnett; 6. Chemical evolution of galaxies and intracluster medium F.Matteucci; 7. Element abundances through the cosmic ages M. Pettini.

Bulk and rare earth abundances in the Luna 16 soil levels A and D.

NASA Technical Reports Server (NTRS)

Gillum, D. E.; Ehmann, W. D.; Wakita, H.; Schmitt, R. A.

1972-01-01

Determination of the abundances of major, minor, and trace elements by means of sequential INAA (instrumental neutron activation analysis) in two Luna 16 soils, at levels A (about 7 cm depth) and D (about 30 cm depth). Abundances of the bulk elements in Luna 16 soils generally agree with the values reported by Vinogradov (1971). Elemental abundances of both bulk and trace elements are nearly the same for the two A and D soil levels. Overall, the chemical compositions of the two Luna 16 soils are more closely related to Apollo 11 soil 10084 than to Apollo 12 and 14 soils, with the exception of TiO2 abundances.-

IRRADIATION METHOD AND APPARATUS

DOEpatents

Cabell, C.P.

1962-12-18

A method and apparatus are described for changing fuel bodies into a process tube of a reactor. According to this method fresh fuel elements are introduced into one end of the tube forcing used fuel elements out the other end. When sufficient fuel has been discharged, a reel and tape arrangement is employed to pull the column of bodies back into the center of the tube. Due provision is made for providing shielding in the tube. (AEC)

Yttrium and rare earth stabilized fast reactor metal fuel

DOEpatents

Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

1992-01-01

To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

49 CFR 501.7 - Administrator's reservations of authority.

Code of Federal Regulations, 2010 CFR

2010-10-01

... under 23 U.S.C. chapter 4, as amended, to: (1) Apportion authorization amounts and distribute obligation... penalties and approve manufacturer fuel economy credit plans under chapter 329. (d) The authority under...

Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

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

Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.

2008-07-15

IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicidemore » was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)« less

Axially staggered seed-blanket reactor-fuel-module construction. [LWBR

DOEpatents

Cowell, G.K.; DiGuiseppe, C.P.

1982-10-28

A heterogeneous nuclear reactor of the seed-blanket type is provided wherein the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements.

Geometric Approach to Orbital Formation Mission Design

DTIC Science & Technology

2004-03-01

limitations, several individuals have used their resources to im- prove upon the Clohessy and Wiltshire model. First order oblateness affects are added to... Clohessy and Wiltshire solutions by Schaub and Alfriend [21] who de- scribe the relative orbit using Delaunay [7] orbital elements. Further perturbation...a methodology using a sliding mode framework. Irvin [11] investigated minimal fuel reconfiguration techniques using the Clohessy and Wiltshire

Influence of the operational parameters on bioelectricity generation in continuous microbial fuel cell, experimental and computational fluid dynamics modelling

NASA Astrophysics Data System (ADS)

Sobieszuk, Paweł; Zamojska-Jaroszewicz, Anna; Makowski, Łukasz

2017-12-01

The influence of the organic loading rate (also known as active anodic chamber volume) on bioelectricity generation in a continuous, two-chamber microbial fuel cell for the treatment of synthetic wastewater, with glucose as the only carbon source, was examined. Ten sets of experiments with different combinations of hydraulic retention times (0.24-1.14 d) and influent chemical oxygen demand concentrations were performed to verify the impact of organic loading rate on the voltage generation capacity of a simple dual-chamber microbial fuel cell working in continuous mode. We found that there is an optimal hydraulic retention time value at which the maximum voltage is generated: 0.41 d. However, there were no similar effects, in terms of voltage generation, when a constant hydraulic retention time with different influent chemical oxygen demand of wastewater was used. The obtained maximal voltage value (600 mV) has also been compared to literature data. Computational fluid dynamics (CFD) was used to calculate the fluid flow and the exit age distribution of fluid elements in the reactor to explain the obtained experimental results and identify the crucial parameters for the design of bioreactors on an industrial scale.

``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

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

Jerde, E.A.; Glasgow, D.C.

1998-09-01

At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate countingmore » leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.« less

PREIRRADIATION MEASUREMENTS OF PIQUA FUEL ELEMENTS NO. P-1111, P-1113, P- 1114, AND P-1120

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

Hubbell, H.J.

1962-11-01

Results of preirradiation measurements and tests performed during the processing and assembly of the individual fuel cylinders contained in Piqua Fuel Elements No. P-1111, P-1113, P-1114, and P-1120 are presented. A description of the techniques and equipment used in obtaining the data is also included. (auth)

Aluminum hydroxide coating thickness measurements and brushing tests on K West Basin fuel elements

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

Pitner, A.L.

1998-09-11

Aluminum hydroxide coating thicknesses were measured on fuel elements stored in aluminum canisters in K West Basin using specially developed eddy current probes . The results were used to estimate coating inventories for MCO fuel,loading. Brushing tests successfully demonstrated the ability to remove the coating if deemed necessary prior to MCO loading.

Photographic combustion characterization of LOX/hydrocarbon type propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1979-01-01

Single element injectors and two fuels were tested with the aim of photographically characterizing observed combustion phenomena. The three injectors tested were the O-F-O triplet, the transverse like on like (TLOL), and the rectangular unlike doublet (RUD). The fuels tested were RP-1 and propane. The hot firings were conducted in a specifically constructed chamber fitted with quartz windows for photographically viewing the impingement spray field. All LOX/HC testing demonstrated coking with the RP-1 fuel leaving far more soot than the propane fuel. No fuel freezing or popping was experienced under the test conditions evaluated. Carbon particle emission and combustion light brilliance increased with Pc for both fuels although RP-1 was far more energetic in this respect. The RSS phenomena appear to be present in the high Pc tests as evidenced by striations in the spray pattern and by separate fuel rich and oxidizer rich areas. The RUD element was also tested as a fuel rich gas generator element by switching the propellant circuits. Excessive sooting occurred at this low mixture ratio (0.55), precluding photographic data.

NEUTRONIC REACTOR CONTROL ELEMENT

DOEpatents

Newson, H.W.

1960-09-13

A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

3D thermal modeling of TRISO fuel coupled with neutronic simulation

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

Hu, Jianwei; Uddin, Rizwan

2010-01-01

The Very High Temperature Gas Reactor (VHTR) is widely considered as one of the top candidates identified in the Next Generation Nuclear Power-plant (NGNP) Technology Roadmap under the U.S . Depanment of Energy's Generation IV program. TRlSO particle is a common element among different VHTR designs and its performance is critical to the safety and reliability of the whole reactor. A TRISO particle experiences complex thermo-mechanical changes during reactor operation in high temperature and high burnup conditions. TRISO fuel performance analysis requires evaluation of these changes on micro scale. Since most of these changes are temperature dependent, 3D thermal modelingmore » of TRISO fuel is a crucial step of the whole analysis package. In this paper, a 3D numerical thermal model was developed to calculate temperature distribution inside TRISO and pebble under different scenarios. 3D simulation is required because pebbles or TRISOs are always subjected to asymmetric thermal conditions since they are randomly packed together. The numerical model was developed using finite difference method and it was benchmarked against ID analytical results and also results reported from literature. Monte-Carlo models were set up to calculate radial power density profile. Complex convective boundary condition was applied on the pebble outer surface. Three reactors were simulated using this model to calculate temperature distribution under different power levels. Two asymmetric boundary conditions were applied to the pebble to test the 3D capabilities. A gas bubble was hypothesized inside the TRISO kernel and 3D simulation was also carried out under this scenario. Intuition-coherent results were obtained and reported in this paper.« less

ELECTROLYTIC SEPARATION PROCESS AND APPARATUS

DOEpatents

McLain, M.E. Jr.; Roberts, M.W.

1962-03-01

A method is given for dissolving stainless steel-c lad fuel elements in dilute acids such as half normal sulfuric acid. The fuel element is made the anode in a Y-shaped electrolytic cell which has a flowing mercury cathode; the stainless steel elements are entrained in the mercury and stripped therefrom by a continuous process. (AEC)

Nuclear breeder reactor fuel element with silicon carbide getter

DOEpatents

Christiansen, David W.; Karnesky, Richard A.

1987-01-01

An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

Synthesizing 2D MoS2 Nanofins on carbon nanospheres as catalyst support for Proton Exchange Membrane Fuel Cells.

PubMed

Hu, Yan; Chua, Daniel H C

2016-06-15

Highly dense 2D MoS2 fin-like nanostructures on carbon nanospheres were fabricated and formed the main catalyst support structure in the oxygen reduction reaction (ORR) for polymer electrolyte membrane (PEM) fuel cells. These nanofins were observed growing perpendicular to the carbon nanosphere surface in random orientations and high resolution transmission electron microscope confirmed 2D layers. The PEM fuel cell test showed enhanced electrochemical activity with good stability, generating over 8.5 W.mgPt(-1) as compared to standard carbon black of 7.4 W.mgPt(-1) under normal operating conditions. Electrochemical Impedance Spectroscopy confirmed that the performance improvement is highly due to the excellent water management of the MoS2 lamellar network, which facilitates water retention at low current density and flood prevention at high current density. Reliability test further demonstrated that these nanofins are highly stable in the electrochemical reaction and is an excellent ORR catalyst support.

Synthesizing 2D MoS2 Nanofins on carbon nanospheres as catalyst support for Proton Exchange Membrane Fuel Cells

PubMed Central

Hu, Yan; Chua, Daniel H. C.

2016-01-01

Highly dense 2D MoS2 fin-like nanostructures on carbon nanospheres were fabricated and formed the main catalyst support structure in the oxygen reduction reaction (ORR) for polymer electrolyte membrane (PEM) fuel cells. These nanofins were observed growing perpendicular to the carbon nanosphere surface in random orientations and high resolution transmission electron microscope confirmed 2D layers. The PEM fuel cell test showed enhanced electrochemical activity with good stability, generating over 8.5 W.mgPt−1 as compared to standard carbon black of 7.4 W.mgPt−1 under normal operating conditions. Electrochemical Impedance Spectroscopy confirmed that the performance improvement is highly due to the excellent water management of the MoS2 lamellar network, which facilitates water retention at low current density and flood prevention at high current density. Reliability test further demonstrated that these nanofins are highly stable in the electrochemical reaction and is an excellent ORR catalyst support. PMID:27302135

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

DOEpatents

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

NASA Technical Reports Server (NTRS)

Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

2012-01-01

With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Prototypical fuel elements mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.

Liquid fuel injection elements for rocket engines

NASA Technical Reports Server (NTRS)

Cox, George B., Jr. (Inventor)

1993-01-01

Thrust chambers for liquid propellant rocket engines include three principal components. One of these components is an injector which contains a plurality of injection elements to meter the flow of propellants at a predetermined rate, and fuel to oxidizer mixture ratio, to introduce the mixture into the combustion chamber, and to cause them to be atomized within the combustion chamber so that even combustion takes place. Evolving from these injectors are tube injectors. These tube injectors have injection elements for injecting the oxidizer into the combustion chamber. The oxidizer and fuel must be metered at predetermined rates and mixture ratios in order to mix them within the combustion chamber so that combustion takes place smoothly and completely. Hence tube injectors are subject to improvement. An injection element for a liquid propellant rocket engine of the bipropellant type is provided which includes tangential fuel metering orifices, and a plurality of oxidizer tube injection elements whose injection tubes are also provided with tangential oxidizer entry slots and internal reed valves.

Development of a Tritium Extruder for ITER Pellet Injection

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

M.J. Gouge; P.W. Fisher

As part of the International Thermonuclear Experimental Reactor (ITER) plasma fueling development program, Oak Ridge National Laboratory (ORNL) has fabricated a pellet injection system to test the mechanical and thermal properties of extruded tritium. Hydrogenic pellets will be used in ITER to sustain the fusion power in the plasma core and may be crucial in reducing first-wall tritium inventories by a process of "isotopic fueling" in which tritium-rich pellets fuel the burning plasma core and deuterium gas fuels the edge. This repeating single-stage pneumatic pellet injector, called the Tritium-Proof-of-Principle Phase II (TPOP-II) Pellet Injector, has a piston-driven mechanical extruder andmore » is designed to extrude and accelerate hydrogenic pellets sized for the ITER device. The TPOP-II program has the following development goals: evaluate the feasibility of extruding tritium and deuterium-tritium (D-T) mixtures for use in future pellet injection systems; determine the mechanical and thermal properties of tritium and D-T extrusions; integrate, test, and evaluate the extruder in a repeating, single-stage light gas gun that is sized for the ITER application (pellet diameter -7 to 8 mm); evaluate options for recycling propellant and extruder exhaust gas; and evaluate operability and reliability of ITER prototypical fueling systems in an environment of significant tritium inventory that requires secondary and room containment systems. In tests with deuterium feed at ORNL, up to 13 pellets per extrusion have been extruded at rates up to 1 Hz and accelerated to speeds of 1.0 to 1.1 km/s, using hydrogen propellant gas at a supply pressure of 65 bar. Initially, deuterium pellets 7.5 mm in diameter and 11 mm in length were produced-the largest cryogenic pellets produced by the fusion program to date. These pellets represent about a 10% density perturbation to ITER. Subsequently, the extruder nozzle was modified to produce pellets that are almost 7.5-mm right circular cylinders. Tritium and D-T pellets have been produced in experiments at the Los Alamos National Laboratory Tritium Systems Test Assembly. About 38 g of tritium have been utilized in the experiment. The tritium was received in eight batches, six from product containers and two from the Isotope Separation System. Two types of runs were made: those in which the material was only extruded and those in which pellets were produced and fired with deuterium propellant. A total of 36 TZ runs and 28 D-T runs have been made. A total of 36 pure tritium runs and 28 D-T mixture runs were made. Extrusion experiments indicate that both T2 and D-T will require higher extrusion forces than D2 by about a factor of two.« less

Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

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

Terrani, Kurt A; Kiggans Jr, James O; McMurray, Jake W

2016-01-01

Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhancemore » heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.« less

FLUID MODERATED REACTOR

DOEpatents

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

1957-10-22

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

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, Norman B.

1998-01-01

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, N.B.

1998-09-08

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

NASA Technical Reports Server (NTRS)

Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

2013-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

Improved gas tagging and cover gas combination for nuclear reactor

DOEpatents

Gross, K.C.; Laug, M.T.

1983-09-26

The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

Review of Rover fuel element protective coating development at Los Alamos

NASA Technical Reports Server (NTRS)

Wallace, Terry C.

1991-01-01

The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

RELAP-7 Development Updates

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

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

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

1963-05-01

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

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

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

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-12-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, eachmore » containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.« less

Development, characterization and dissolution behavior of calcium-aluminoborate glass wasteforms to immobilize rare-earth oxides.

PubMed

Kim, Miae; Corkhill, Claire L; Hyatt, Neil C; Heo, Jong

2018-03-28

Calcium-aluminoborate (CAB) glasses were developed to sequester new waste compositions made of several rare-earth oxides generated from the pyrochemical reprocessing of spent nuclear fuel. Several important wasteform properties such as waste loading, processability and chemical durability were evaluated. The maximum waste loading of the CAB compositions was determined to be ~56.8 wt%. Viscosity and the electrical conductivity of the CAB melt at 1300 °C were 7.817 Pa·s and 0.4603 S/cm, respectively, which satisfies the conditions for commercial cold-crucible induction melting (CCIM) process. Addition of rare-earth oxides to CAB glasses resulted in dramatic decreases in the elemental releases of B and Ca in aqueous dissolution experiments. Normalized elemental releases from product consistency standard chemical durability test were <3.62·10 -5  g·m -2 for Nd, 0.009 g·m -2 for Al, 0.067 g·m -2 for B and 0.073 g·m -2 for Ca (at 90, after 7 days, for SA/V = 2000m -1 ); all meet European and US regulation limits. After 20 d of dissolution, a hydrated alteration layer of ~ 200-nm-thick, Ca-depleted and Nd-rich, was formed at the surface of CAB glasses with 20 mol% Nd 2 O 3 whereas boehmite [AlO(OH)] secondary crystalline phases were formed in pure CAB glass that contained no Nd 2 O 3 .

Risk Stratification of Stress Fractures and Prediction of Return-to-Duty

DTIC Science & Technology

2016-12-01

African-American) and 20 male; [month 7-28] Done 4) Perform standard whole bone finite element analysis [month 7-28]. Done 5) Perform data cleaning and...NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Kristin Popp M Mary Betty Diamond 5d. PROJECT NUMBER Mary Bouxsein 5e. TASK NUMBER E-Mail: kpopp

40 CFR 86.113-94 - Fuel specifications.

Code of Federal Regulations, 2013 CFR

2013-07-01

... (CONTINUED) CONTROL OF EMISSIONS FROM NEW AND IN-USE HIGHWAY VEHICLES AND ENGINES Emission Regulations for.... Value Octane, Research, Min D2699 93 Sensitivity, Min 7.5 Lead (organic): g/U.S. gal. (g/liter) D3237 1... accumulation. (ii) The octane rating of the gasoline used shall be no higher than 1.0 Research octane number...

Method and apparatus for diagnosing breached fuel elements

DOEpatents

Gross, K.C.; Lambert, J.D.B.; Nomura, S.

1987-03-02

The invention provides an apparatus and method for diagnosing breached fuel elements in a nuclear reactor. A detection system measures the activity of isotopes from the cover gas in the reactor. A data acquisition and processing system monitors the detection system and corrects for the effects of the cover-gas clean up system on the measured activity and further calculates the derivative curve of the corrected activity as a function of time. A plotting system graphs the derivative curve, which represents the instantaneous release rate of fission gas from a breached fuel element. 8 figs.

JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

DOEpatents

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

1959-05-12

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

40 CFR 1065.703 - Distillate diesel fuel.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 1 Cetane Number 40-50 40-50 40-50 ASTM D613-05. Distillation range: Initial boiling point °C 171-204 171-204 171-204 ASTM D86-07a. 10 pct. point 204-238 204-238 204-238 50 pct. point 243-282 243-282 243...-37 ASTM D4052-96e01. Total sulfur, ultra low sulfur mg/kg 7-15 See 40 CFR 80.580. Total sulfur, low...

Uranium from German Nuclear Power Projects of the 1940s— A Nuclear Forensic Investigation

PubMed Central

Mayer, Klaus; Wallenius, Maria; Lützenkirchen, Klaus; Horta, Joan; Nicholl, Adrian; Rasmussen, Gert; van Belle, Pieter; Varga, Zsolt; Buda, Razvan; Erdmann, Nicole; Kratz, Jens-Volker; Trautmann, Norbert; Fifield, L Keith; Tims, Stephen G; Fröhlich, Michaela B; Steier, Peter

2015-01-01

Here we present a nuclear forensic study of uranium from German nuclear projects which used different geometries of metallic uranium fuel.3b,d, 4 Through measurement of the 230Th/234U ratio, we could determine that the material had been produced in the period from 1940 to 1943. To determine the geographical origin of the uranium, the rare-earth-element content and the 87Sr/86Sr ratio were measured. The results provide evidence that the uranium was mined in the Czech Republic. Trace amounts of 236U and 239Pu were detected at the level of their natural abundance, which indicates that the uranium fuel was not exposed to any major neutron fluence. PMID:26501922

Enrichment Zoning Options for the Small Nuclear Rocket Engine (SNRE)

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

Bruce G. Schnitzler; Stanley K. Borowski

2010-07-01

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. In NASA’s recent Mars Design Reference Architecture (DRA) 5.0 study (NASA-SP-2009-566, July 2009), nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option because of its high thrust and high specific impulse (-900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. An extensive nuclear thermal rocket technology development effortmore » was conducted from 1955-1973 under the Rover/NERVA Program. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art design incorporating lessons learned from the very successful technology development program. Past activities at the NASA Glenn Research Center have included development of highly detailed MCNP Monte Carlo transport models of the SNRE and other small engine designs. Preliminary core configurations typically employ fuel elements with fixed fuel composition and fissile material enrichment. Uniform fuel loadings result in undesirable radial power and temperature profiles in the engines. Engine performance can be improved by some combination of propellant flow control at the fuel element level and by varying the fuel composition. Enrichment zoning at the fuel element level with lower enrichments in the higher power elements at the core center and on the core periphery is particularly effective. Power flattening by enrichment zoning typically results in more uniform propellant exit temperatures and improved engine performance. For the SNRE, element enrichment zoning provided very flat radial power profiles with 551 of the 564 fuel elements within 1% of the average element power. Results for this and alternate enrichment zoning options for the SNRE are compared.« less

Non-Catalytic Reforming with Applications to Portable Power

DTIC Science & Technology

2013-10-01

and J.J. Beaman, Jr., “Freeform Fabrication of Non-Metallic Objects by Selective Laser Sintering and Infiltration”, Materials Science Forum, 561-565...for syngas production from jet fuel using various methods including catalysts [4, 47-54] and plasmas [55]. Investigations of noncatalytic reforming...Combustion of n-butanol in a spark -ignition IC engine. Fuel. 89(7): p. 1573-1582. 32. Behrens, D.A., I.C. Lee, and C.M. Waits, Catalytic combustion of

DoD Contractors’ Safety Manual for Ammunition and Explosives

DTIC Science & Technology

1986-03-01

Trifluoride ClF ------------------------- II A3 Hydrogen Peroxide Greatel than 52% H202 II3 A Liquid Fluorine LF 2...7.3 68 Fluorine (Liquid) 12.6 -306 Furfuryl Alcohol 9.4 68 Hydrocarbon Fuel JP-4 6.35 60 Hydrocarbon Fuel JP-5 6.84 60 Hydrogen Peroxide (90%) 11.6...specifically authorized as process requirements. 2. Some flammable liquids such as paints, varnishes , and enamels may, under certain conditions, be

FUEL ELEMENT CONSTRUCTION

DOEpatents

Zumwalt, L.R.

1961-08-01

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

NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

DOEpatents

Brooks, H.

1960-04-26

A description is given for a fuel element comprising a body of uranium metal or an uranium compound dispersed in a matrix material made from magnesium, calcium, or barium and a stainless steel jacket enclosing the body.

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

Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

A Multi-Methods Approach to HRA and Human Performance Modeling: A Field Assessment

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

Jacques Hugo; David I Gertman

2012-06-01

The Advanced Test Reactor (ATR) is a research reactor at the Idaho National Laboratory is primarily designed and used to test materials to be used in other, larger-scale and prototype reactors. The reactor offers various specialized systems and allows certain experiments to be run at their own temperature and pressure. The ATR Canal temporarily stores completed experiments and used fuel. It also has facilities to conduct underwater operations such as experiment examination or removal. In reviewing the ATR safety basis, a number of concerns were identified involving the ATR canal. A brief study identified ergonomic issues involving the manual handlingmore » of fuel elements in the canal that may increase the probability of human error and possible unwanted acute physical outcomes to the operator. In response to this concern, that refined the previous HRA scoping analysis by determining the probability of the inadvertent exposure of a fuel element to the air during fuel movement and inspection was conducted. The HRA analysis employed the SPAR-H method and was supplemented by information gained from a detailed analysis of the fuel inspection and transfer tasks. This latter analysis included ergonomics, work cycles, task duration, and workload imposed by tool and workplace characteristics, personal protective clothing, and operational practices that have the potential to increase physical and mental workload. Part of this analysis consisted of NASA-TLX analyses, combined with operational sequence analysis, computational human performance analysis (CHPA), and 3D graphical modeling to determine task failures and precursors to such failures that have safety implications. Experience in applying multiple analysis techniques in support of HRA methods is discussed.« less

ADS Model in the TIRELIRE-STRATEGIE Fuel Cycle Simulation Code Application to Minor Actinides Transmutation Studies

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

Garzenne, Claude; Massara, Simone; Tetart, Philippe

2006-07-01

Accelerator Driven Systems offer the advantage, thanks to the core sub-criticality, to burn highly radioactive elements such as americium and curium in a dedicated stratum, and then to avoid polluting with these elements the main part of the nuclear fleet, which is optimized for electricity production. This paper presents firstly the ADS model implemented in the fuel cycle simulation code TIRELIRE-STRATEGIE that we developed at EDF R and D Division for nuclear power scenario studies. Then we show and comment the results of TIRELIRE-STRATEGIE calculation of a transition scenario between the current French nuclear fleet, and a fast reactor fleetmore » entirely deployed towards the end of the 21. century, consistently with the EDF prospective view, with 3 options for the minor actinides management:1) vitrified with fission products to be sent to the final disposal; 2) extracted together with plutonium from the spent fuel to be transmuted in Generation IV fast reactors; 3) eventually extracted separately from plutonium to be incinerated in a ADSs double stratum. The comparison of nuclear fuel cycle material fluxes and inventories between these options shows that ADSs are not more efficient than critical fast reactors for reducing the high level waste radio-toxicity; that minor actinides inventory and fluxes in the fuel cycle are more than twice as high in case of a double ADSs stratum than in case of minor actinides transmutation in Generation IV FBRs; and that about fourteen 400 MWth ADS are necessary to incinerate minor actinides issued from a 60 GWe Generation IV fast reactor fleet, corresponding to the current French nuclear fleet installed power. (authors)« less

Ion chromatographic determination of sulfur in fuels

NASA Technical Reports Server (NTRS)

Mizisin, C. S.; Kuivinen, D. E.; Otterson, D. A.

1978-01-01

The sulfur content of fuels was determined using an ion chromatograph to measure the sulfate produced by a modified Parr bomb oxidation. Standard Reference Materials from the National Bureau of Standards, of approximately 0.2 + or - 0.004% sulfur, were analyzed resulting in a standard deviation no greater than 0.008. The ion chromatographic method can be applied to conventional fuels as well as shale-oil derived fuels. Other acid forming elements, such as fluorine, chlorine and nitrogen could be determined at the same time, provided that these elements have reached a suitable ionic state during the oxidation of the fuel.

DART model for irradiation-induced swelling of uranium silicide dispersion fuel elements

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

Rest, J.; Hofman, G.L.

1999-04-01

Models for the interaction of uranium silicide dispersion fuels with an aluminum matrix, for the resultant reaction product swelling, and for the calculation of the stress gradient within the fuel particles are described within the context of DART fission-gas-induced swelling models. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by comparing DART calculations with irradiation data for the swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al in variously designed dispersion fuel elements.

Alternative Fuels. Hearing before the Subcommittee on Energy and Power of the Committee on Energy and Commerce, House of Representatives, One Hundred Third Congress, First Session, October 7, 1993

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

Not Available

1994-01-01

The 1990 Clean Air Act Amendments, in requiring the States to reduce air pollution, resulted in new regulations on gasoline composition, on automobile emissions and mandates on the use of alternative fuels. The 1992 Energy Policy Act includes provisions aiming to reduce energy dependence by increasing the use of alternative, non-oil fuels in certain vehicle fleets. This hearing focus on where development of alternative fuels stands today. Testimony is presented by the following: T. Jorling, NY State Dept. of Environmental Conservation; R.L. Klimisch, American Automobile Manufactures ASS., D. Smith, Chevron USA Products Co., and P. Wuebben, clean fuels officer, Southmore » Coast Air quality Management District.« less

Introducing the global carbon cycle to middle school students with a 14C research project

NASA Astrophysics Data System (ADS)

Brodman Larson, L.; Phillips, C. L.; LaFranchi, B. W.

2012-12-01

Global Climate Change (GCC) is currently not part of the California Science Standards for 7th grade. Required course elements, however, such as the carbon cycle, photosynthesis, and cellular respiration could be linked to global climate change. Here we present a lesson plan developed in collaboration with scientists from Lawrence Livermore National Laboratory, to involve 7th grade students in monitoring of fossil fuel emissions in the Richmond/San Pablo area of California. -The lesson plan is a Greenhouse Gas/Global Climate Change Unit, with an embedded research project in which students will collect plant samples from various locals for analysis of 14C, to determine if there is a correlation between location and how much CO2 is coming from fossil fuel combustion. Main learning objectives are for students to: 1) understand how fossil fuel emissions impact the global carbon cycle, 2) understand how scientists estimate fossil CO2 emissions, and 3) engage in hypothesis development and testing. This project also engages students in active science learning and helps to develop responsibility, two key factors for adolescentsWe expect to see a correlation between proximity to freeways and levels of fossil fuel emissions. This unit will introduce important GCC concepts to students at a younger age, and increase their knowledge about fossil fuel emissions in their local environment, as well as the regional and global impacts of fossil emissions.

HEAT TREATED U-Nb ALLOYS

DOEpatents

McGeary, R.K.; Justusson, W.M.

1959-11-24

A fuel element for a nuclear reactor is described comprising an alloy containing uranium and from 7 to 20 wt.% niobium, the alloy being substantially in the gamma phase and having been produced by working an ingot of the alloy into the desired shape, homogenizing it by annealing it at a temperature in the gamma phase field, and quenching it to retain the gamma phase structure of the alloy.

Prime Contract Awards Alphabetically by Contractor, By State or Country, and Place, FY 84. Part 9. (Potomac Edison Company - Southern Alcohol Fuels, Inc).

DTIC Science & Technology

1984-01-01

NO0O00 CD (o- V NOOOOOOOOON- No if)0 NY- 11) tD VNC O 0 ’ n’m - - -NrN VO0 K) 0 -0-- C) - m)- 00 1 - (D NO 0 - fC.) q.0000-N 0- 0 -000000000ow 00 00 0...C’) < <’ C’) 0 0M0 0 -N’q7- n7 - 0) 0 0o t (D 0 N 0 0 N. < 0_.00 ()1OOOC) 0(D0 20 -10 -Ui) 2j- a20 2(0( tD jU) -1 7 210 21)IN CL< . C ( 0.0 D00 <- N...N Nq NDI) 07 ( I C 0---- 0 0- if -C)- -) if - tD - 0 N ~ CD 0 0 0 0 CI) Nl M) 0) .0 C) 0) V) C).- V. - 0).- Ur) -- 0 - q- (D- ------------ 0) - 9(0 (D0

Advances in cryogenic foam insulations.

NASA Technical Reports Server (NTRS)

Lemons, C. R.; Salmassy, O. K.; Watts, C. R.

1971-01-01

Description of a discretely oriented thread-reinforced polyurethane foam thermal insulation system for liquid hydrogen fuel tanks. The 3-D foam and glass liner composite is designed to be adhesively bonded to the inside surface of the tank wall and to be in direct contact with liquid hydrogen. All elements of this insulation composite are capable of sustaining the loads and environmental conditions imposed by testing under simulated Space Shuttle vehicle requirements at temperatures between -423 and +350 F.

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

Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searchedmore » and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.« less

Chemical characterization of the fine particle emissions from commercial aircraft engines during the Aircraft Particle Emissions eXperiment (APEX) 1 to 3.

PubMed

Kinsey, J S; Hays, M D; Dong, Y; Williams, D C; Logan, R

2011-04-15

This paper addresses the need for detailed chemical information on the fine particulate matter (PM) generated by commercial aviation engines. The exhaust plumes of seven turbofan engine models were sampled as part of the three test campaigns of the Aircraft Particle Emissions eXperiment (APEX). In these experiments, continuous measurements of black carbon (BC) and particle surface-bound polycyclic aromatic compounds (PAHs) were conducted. In addition, time-integrated sampling was performed for bulk elemental composition, water-soluble ions, organic and elemental carbon (OC and EC), and trace semivolatile organic compounds (SVOCs). The continuous BC and PAH monitoring showed a characteristic U-shaped curve of the emission index (EI or mass of pollutant/mass of fuel burned) vs fuel flow for the turbofan engines tested. The time-integrated EIs for both elemental composition and water-soluble ions were heavily dominated by sulfur and SO(4)(2-), respectively, with a ∼2.4% median conversion of fuel S(IV) to particle S(VI). The corrected OC and EC emission indices obtained in this study ranged from 37 to 83 mg/kg and 21 to 275 mg/kg, respectively, with the EC/OC ratio ranging from ∼0.3 to 7 depending on engine type and test conditions. Finally, the particle SVOC EIs varied by as much as 2 orders of magnitude with distinct variations in chemical composition observed for different engine types and operating conditions.

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

Tree bark suber-included particles: A long-term accumulation site for elements of atmospheric origin

NASA Astrophysics Data System (ADS)

Catinon, Mickaël; Ayrault, Sophie; Spadini, Lorenzo; Boudouma, Omar; Asta, Juliette; Tissut, Michel; Ravanel, Patrick

2011-02-01

The deposition of atmospheric elements on and into the bark of 4-year-old Fraxinus excelsior L. was studied. The elemental composition of the suber tissue was established through ICP-MS analysis and the presence of solid mineral particles included in this suber was established and described through SEM-EDX. Fractionation of the suber elements mixture was obtained after ashing at 550 °C through successive water (C fraction) and HNO 3 2 M (D fraction) extraction, leading to an insoluble residue mainly composed of the solid mineral particles (E fraction). The triplicated % weight of C, D and E were respectively 34.4 ± 2.7, 64.8 ± 2.7 and 0.8 ± 0.1% of the suber ashes weight. The main component of C was K, of D was Ca. Noticeable amounts of Mg were also observed in D. The E fraction, composed of insoluble particles, was mostly constituted of geogenic products, with elements such as Si, Al, K, Mg, representing primary minerals. E also contained Ca 3(PO 4) 2 and concentrated the main part of Pb and Fe. Moreover, The SEM-EDX analysis evidenced that this fraction also concentrated several types of fly ashes of industrial origin. The study of the distribution between C, D and E was analysed through ICP-MS with respect to their origin. The origin of the elements found in such bark was either geogenic (clay, micas, quartz…), anthropogenic or biogenic (for instance large amounts of solid Ca organic salts having a storage role). As opposed to the E fraction, the C fraction, mainly composed of highly soluble K+ is characteristic of a biological pool of plant origin. In fraction D, the very high amount of Ca++ corresponds to two different origins: biological or acid soluble minerals such as calcite. Furthermore, the D fraction contains the most part of pollutants of anthropic origin such as Zn, Cu, Ni, Co, Cd. As a whole, the fractionation procedure of the suber samples allows to separate elements as a function of their origin but also gives valuable information on distribution and speciation of trace elements.

Methods for making a porous nuclear fuel element

DOEpatents

Youchison, Dennis L; Williams, Brian E; Benander, Robert E

2014-12-30

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Characterization of chemical and particulate emissions from aircraft engines

NASA Astrophysics Data System (ADS)

Agrawal, Harshit; Sawant, Aniket A.; Jansen, Karel; Wayne Miller, J.; Cocker, David R.

2008-06-01

This paper presents a series of measurements from four on-wing, commercial aircraft engines, including two newer CFM56-7 engines and two earlier CFM56-3 engines. Samples were collected from each engine using a probe positioned behind the exhaust nozzle of the aircraft, chocked on a concrete testing pad. The emission factors for particulate matter mass, elemental and organic carbon, carbonyls, polycyclic aromatic hydrocarbons, n-alkanes, dioxins, metals and ions are reported for four different engine power setting modes. The emissions indices of particulate matter, elemental and organic carbon are highly power dependent for these engines. Particulate matter emission indices (g kg-1 fuel) are found to increase from 1.1E-02 to 2.05E-01 with increase in power from idle to 85%. The elemental carbon to organic carbon varies from 0.5 to 3.8 with change in power from idle to 85%. The carbonyl emissions are dominated by formaldehyde. The emission index of formaldehyde ranges from 2.3E-01 to 4.8E-01 g kg-1 fuel. The distribution of metals depends on the difference in the various engines. The dioxin emissions from the aircraft engines are observed to be below detection limit.

Gaseous Surrogate Hydrocarbons for a Hifire Scramjet that Mimic Opposed Jet Extinction Limits for Cracked JP Fuels

NASA Technical Reports Server (NTRS)

Pellett, Gerald L.; Vaden, Sarah N.; Wilson, Lloyd G.

2008-01-01

This paper describes, first, the top-down methodology used to define simple gaseous surrogate hydrocarbon (HC) fuel mixtures for a hypersonic scramjet combustion subtask of the HiFIRE program. It then presents new and updated Opposed Jet Burner (OJB) extinction-limit Flame Strength (FS) data obtained from laminar non-premixed HC vs. air counterflow diffusion flames at 1-atm, which follow from earlier investigations. FS represents a strain-induced extinction limit based on cross-section-average air jet velocity, U(sub air), that sustains combustion of a counter jet of gaseous fuel just before extinction. FS uniquely characterizes a kinetically limited fuel combustion rate. More generally, Applied Stress Rates (ASRs) at extinction (U(sub air) normalized by nozzle or tube diameter, D(sub n or t) can directly be compared with extinction limits determined numerically using either a 1-D or (preferably) a 2-D Navier Stokes simulation with detailed transport and finite rate chemistry. The FS results help to characterize and define three candidate surrogate HC fuel mixtures that exhibit a common FS 70% greater than for vaporized JP-7 fuel. These include a binary fuel mixture of 64% ethylene + 36% methane, which is our primary recommendation. It is intended to mimic the critical flameholding limit of a thermally- or catalytically-cracked JP-7 like fuel in HiFIRE scramjet combustion tests. Our supporting experimental results include: (1) An idealized kinetically-limited ASR reactivity scale, which represents maximum strength non-premixed flames for several gaseous and vaporized liquid HCs; (2) FS characterizations of Colket and Spadaccini s suggested ternary surrogate, of 60% ethylene + 30% methane + 10% n-heptane, which matches the ignition delay of a typical cracked JP fuel; (3) Data showing how our recommended binary surrogate, of 64% ethylene + 36% methane, has an identical FS; (4) Data that characterize an alternate surrogate of 44% ethylene + 56% ethane with identical FS and nearly equal molecular weights; this could be useful when systematically varying the fuel composition. However, the mixture liquefies at much lower pressure, which limits on-board storage of gaseous fuel; (5) Dynamic Flame Weakening results that show how oscillations in OJB input flow (and composition) can weaken (extinguish) surrogate flames up to 200 Hz, but the weakening is 2.5x smaller compared to pure methane; and finally, (6) FS limits at 1-atm that compare with three published 1-D numerical OJB extinction results using four chemical kinetic models. The methane kinetics generally agree closely at 1-atm, whereas, the various ethylene models predict extinction limits that average 45% high, which represents a significant problem for numerical simulation of surrogate-based flameholding in a scramjet cavity. Finally, we continue advocating the FS approach as more direct and fundamental for assessing idealized scramjet flameholding potentials than measurements of "unstrained" premixed laminar burning velocity or blowout in a Perfectly Stirred Reactor.

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 New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

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

Senor, David J.; Painter, Chad L.; Geelhood, Ken J.

2007-12-01

Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling,more » core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.« less

Apollo 12 Mission image - Alan Bean unloads ALSEP RTG fuel element

NASA Image and Video Library

1969-11-19

AS12-46-6790 (19 Nov. 1969) --- Astronaut Alan L. Bean, lunar module pilot, is photographed at quadrant II of the Lunar Module (LM) during the first Apollo 12 extravehicular activity (EVA) on the moon. This picture was taken by astronaut Charles Conrad Jr., commander. Here, Bean is using a fuel transfer tool to remove the fuel element from the fuel cask mounted on the LM's descent stage. The fuel element was then placed in the Radioisotope Thermoelectric Generator (RTG), the power source for the Apollo Lunar Surface Experiments Package (ALSEP) which was deployed on the moon by the two astronauts. The RTG is next to Bean's right leg. While astronauts Conrad and Bean descended in the LM "Intrepid" to explore the Ocean of Storms region of the moon, astronaut Richard F. Gordon Jr., command module pilot, remained with the Command and Service Modules (CSM) "Yankee Clipper" in lunar orbit.

Reliability analysis of dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

2008-03-01

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

Numerical Optimisation in Non Reacting Conditions of the Injector Geometry for a Continuous Detonation Wave Rocket Engine

NASA Astrophysics Data System (ADS)

Gaillard, T.; Davidenko, D.; Dupoirieux, F.

2015-06-01

The paper presents the methodology and the results of a numerical study, which is aimed at the investigation and optimisation of different means of fuel and oxidizer injection adapted to rocket engines operating in the rotating detonation mode. As the simulations are achieved at the local scale of a single injection element, only one periodic pattern of the whole geometry can be calculated so that the travelling detonation waves and the associated chemical reactions can not be taken into account. Here, separate injection of fuel and oxidizer is considered because premixed injection is handicapped by the risk of upstream propagation of the detonation wave. Different associations of geometrical periodicity and symmetry are investigated for the injection elements distributed over the injector head. To analyse the injection and mixing processes, a nonreacting 3D flow is simulated using the LES approach. Performance of the studied configurations is analysed using the results on instantaneous and mean flowfields as well as by comparing the mixing efficiency and the total pressure recovery evaluated for different configurations.

Nano-Zirconium Tungstate Reinforced Liquid Crystalline Thermosetting Composites with Near Zero Thermal Expansion

DTIC Science & Technology

2015-06-25

5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER 7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) Washington State University, Pullman...1234. 5c. PROGRAM ELEMENT NUMBER. Enter all program element numbers as they appear in the report, e.g. 61101A. 5d. PROJECT NUMBER. Enter all... project numbers as they appear in the report, e.g. 1F665702D1257; ILIR. 5e. TASK NUMBER. Enter all task numbers as they appear in the report

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.

Boiler Briquette Coal versus Raw Coal: Part I-Stack Gas Emissions.

PubMed

Ge, Su; Bai, Zhipeng; Liu, Weili; Zhu, Tan; Wang, Tongjian; Qing, Sheng; Zhang, Junfeng

2001-04-01

Stack gas emissions were characterized for a steam-generating boiler commonly used in China. The boiler was tested when fired with a newly formulated boiler briquette coal (BB-coal) and when fired with conventional raw coal (R-coal). The stack gas emissions were analyzed to determine emission rates and emission factors and to develop chemical source profiles. A dilution source sampling system was used to collect PM on both Teflon membrane filters and quartz fiber filters. The Teflon filters were analyzed gravimetrically for PM 10 and PM 2.5 mass concentrations and by X-ray fluorescence (XRF) for trace elements. The quartz fiber filters were analyzed for organic carbon (OC) and elemental carbon (EC) using a thermal/optical reflectance technique. Sulfur dioxide was measured using the standard wet chemistry method. Carbon monoxide was measured using an Orsat combustion analyzer. The emission rates of the R-coal combustion (in kg/hr), determined using the measured stack gas concentrations and the stack gas emission rates, were 0.74 for PM 10 , 0.38 for PM 25 , 20.7 for SO 2 , and 6.8 for CO, while those of the BB-coal combustion were 0.95 for PM 10 , 0.30 for PM 2 5 , 7.5 for SO 2 , and 5.3 for CO. The fuel-mass-based emission factors (in g/kg) of the R-coal, determined using the emission rates and the fuel burn rates, were 1.68 for PM 10 , 0.87 for PM 25 , 46.7 for SO 2 , and 15 for CO, while those of the BB-coal were 2.51 for PM 10 , 0.79 for PM 2.5 , 19.9 for SO 2 , and 14 for CO. The task-based emission factors (in g/ton steam generated) of the R-coal, determined using the fuel-mass-based emission factors and the coal/ steam conversion factors, were 0.23 for PM 10 , 0.12 for PM 2.5 , 6.4 for SO 2 , and 2.0 for CO, while those of the BB-coal were 0.30 for PM 10 , 0.094 for PM 2.5 , 2.4 for SO 2 , and 1.7 for CO. PM 10 and PM 2.5 elemental compositions are also presented for both types of coal tested in the study.

Boiler briquette coal versus raw coal: Part I--Stack gas emissions.

PubMed

Ge, S; Bai, Z; Liu, W; Zhu, T; Wang, T; Qing, S; Zhang, J

2001-04-01

Stack gas emissions were characterized for a steam-generating boiler commonly used in China. The boiler was tested when fired with a newly formulated boiler briquette coal (BB-coal) and when fired with conventional raw coal (R-coal). The stack gas emissions were analyzed to determine emission rates and emission factors and to develop chemical source profiles. A dilution source sampling system was used to collect PM on both Teflon membrane filters and quartz fiber filters. The Teflon filters were analyzed gravimetrically for PM10 and PM2.5 mass concentrations and by X-ray fluorescence (XRF) for trace elements. The quartz fiber filters were analyzed for organic carbon (OC) and elemental carbon (EC) using a thermal/optical reflectance technique. Sulfur dioxide was measured using the standard wet chemistry method. Carbon monoxide was measured using an Orsat combustion analyzer. The emission rates of the R-coal combustion (in kg/hr), determined using the measured stack gas concentrations and the stack gas emission rates, were 0.74 for PM10, 0.38 for PM2.5, 20.7 for SO2, and 6.8 for CO, while those of the BB-coal combustion were 0.95 for PM10, 0.30 for PM2.5, 7.5 for SO2, and 5.3 for CO. The fuel-mass-based emission factors (in g/kg) of the R-coal, determined using the emission rates and the fuel burn rates, were 1.68 for PM10, 0.87 for PM2.5, 46.7 for SO2, and 15 for CO, while those of the BB-coal were 2.51 for PM10, 0.79 for PM2.5, 19.9 for SO2, and 14 for CO. The task-based emission factors (in g/ton steam generated) of the R-coal, determined using the fuel-mass-based emission factors and the coal/steam conversion factors, were 0.23 for PM10, 0.12 for PM2.5, 6.4 for SO2, and 2.0 for CO, while those of the BB-coal were 0.30 for PM10, 0.094 for PM2.5, 2.4 for SO2, and 1.7 for CO. PM10 and PM2.5 elemental compositions are also presented for both types of coal tested in the study.

Dissolution flowsheet for high flux isotope reactor fuel

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

Foster, T.

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3O 8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2. The HFIR fuel cores will be dissolved and the recovered U will be down-blendedmore » into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg concentration.« less

Use of Life Cycle Costing in the Development of Standards

DTIC Science & Technology

1988-12-01

Bringhurst, Mountain Fuel Supply 6. T. Joseph Cardenas, McGraw Hill 7. Sally A. Hooks, EEI 8. Merle F. McBride, Owens - Corning Fiberglas 9. Kenneth D...Stanonik, GAMA 23. Clifford D. Smith, Owens - Corning Fiberglas 24. Chris Thomaidis, Department of Housing & Urban Development 25. Adrian Tuluca, Steven...Professionals, McGraw-Hill Book Company, 1981. 8. Stewart, Rodney D., Cost EstimatinQ, John Wiley & Sons, Inc., 1982. 9. Owens Corning , Telephone

Impact of anaerobic oxidation of methane on the geochemical cycle of redox-sensitive elements at cold-seep sites of the northern South China Sea

NASA Astrophysics Data System (ADS)

Hu, Yu; Feng, Dong; Liang, Qianyong; Xia, Zhen; Chen, Linying; Chen, Duofu

2015-12-01

Cold hydrocarbon seepage is a frequently observed phenomenon along continental margins worldwide. However, little is known about the impact of seeping fluids on the geochemical cycle of redox-sensitive elements. Pore waters from four gravity cores (D-8, D-5, D-7, and D-F) collected from cold-seep sites of the northern South China Sea were analyzed for SO42-, Mg2+, Ca2+, Sr2+, dissolved inorganic carbon (DIC), δ13CDIC, dissolved Fe, Mn, and trace elements (e.g. Mo, U). The sulfate concentration-depth profiles, δ13CDIC values and (ΔDIC+ΔCa2++ΔMg2+)/ΔSO42- ratios suggest that organoclastic sulfate reduction (OSR) is the dominant process in D-8 core. Besides OSR, anaerobic oxidation of methane (AOM) is partially responsible for depletion of sulfate at D-5 and D-7 cores. The sulfate consumption at D-F core is predominantly caused by AOM. The depth of sulfate-methane interface (SMI) and methane diffusive flux of D-F core are calculated to be ~7 m and 0.035 mol m-2 yr-1, respectively. The relatively shallow SMI and high methane flux at D-F core suggest the activity of gas seepage in this region. The concentrations of dissolved uranium (U) were inferred to decrease significantly within the iron reduction zone. It seems that AOM has limited influence on the U geochemical cycling. In contrast, a good correlation between the consumption of sulfate and the removal of molybdenum (Mo) suggests that AOM has a significantly influence on the geochemical cycle of Mo at cold seeps. Accordingly, cold seep environments may serve as an important potential sink in the marine geochemical cycle of Mo.

Combined catalysts for the combustion of fuel in gas turbines

DOEpatents

Anoshkina, Elvira V.; Laster, Walter R.

2012-11-13

A catalytic oxidation module for a catalytic combustor of a gas turbine engine is provided. The catalytic oxidation module comprises a plurality of spaced apart catalytic elements for receiving a fuel-air mixture over a surface of the catalytic elements. The plurality of catalytic elements includes at least one primary catalytic element comprising a monometallic catalyst and secondary catalytic elements adjacent the primary catalytic element comprising a multi-component catalyst. Ignition of the monometallic catalyst of the primary catalytic element is effective to rapidly increase a temperature within the catalytic oxidation module to a degree sufficient to ignite the multi-component catalyst.

Database Design for Structural Analysis and Design Optimization.

DTIC Science & Technology

1984-10-01

2) . Element number of nodes IELT NPAR(2) " Stress printing flag IPST NPAR(2) Element material angle BETA NPAR(2) Element thickness THICK NPAR(2...number LM 3*NPAR(17)*NPAR(2) Element nodal coordinates XYZ 3*NPAR(17)*NPAR(2) Element number of nodes IELT NPAR(2) Element geometry number of nodes IELTX...D.O.F. number LM 6*NPAR(7)*NPAR(2) Element number of nodes IELT NPAR(2) Material property set number MATP NPAR(2) Material constants PROP NPAR(17

THE FUEL ELEMENT GRAPHITE. Project DRAGON.

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

Graham, L.W.; Price, M.S.T.

1963-01-15

The main requirements of a fuel element graphite for reactors based on the Dragon concept are low transmission coefficient for fission products, dimensional stability under service conditions, high strength, high thermal conductivity, high purity, and high resistance to oxidation. Since conclusions reached in early 1960, a considerable amount of information has accumulated concerning the likely behaviour of graphites in high temperature reactor systems, particularly data on dimensional stability under irradiation. The influence of this new knowledge on the development of fuel element graphite with the Dragon Project is discussed in detail in the final section of this paper.

Low exchange element for nuclear reactor

DOEpatents

Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

1985-01-01

A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

SEM in situ MiniCantilever Beam Bending of U-10Mo/Zr/Al Fuel Elements

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

Mook, William; Baldwin, Jon K.; Martinez, Ricardo M.

2014-06-16

In this work, the fracture behavior of Al/Zr and Zr/dU-10Mo interfaces was measured via the minicantilever bend technique. The energy dissipation rates were found to be approximately 3.7-5 mj/mm 2 and 5.9 mj/mm 2 for each interface, respectively. It was found that in order to test the Zr/U-10Mo interface, location of the hinge of the cantilever was a key parameter. While this test could be adapted to hot cell use through careful alignment fixturing and measurement of crack lengths with an optical microscope (as opposed to SEM, which was used here out of convenience), machining of the cantilevers via MiniMillmore » in such a way as to locate the interfaces at the cantilever hinge, as well as proper placement of a femtosecond laser notch will continue to be key challenges in a hot cell environment.« less

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit corrosion of certain potential NTR fuel forms. Additional diagnostic upgrades included in the present NTREES set up include the addition of a gamma ray spectrometer located near the vent filter to detect uranium fuel particles exiting the fuel element in the propellant exhaust stream to provide additional information any material loss occurring during testing. Other aspects of the upgrade included reworking NTREES to reduce the operational complexity of the system despite the increased complexity of the induction heating system. To this end, many of the controls were consolidated on fewer panels. As part of this upgrade activity, the Safety Assessment (SA) and the Standard Operating Procedures (SOPs) for NTREES were extensively rewritten. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can be accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements.

NUCLEAR REACTOR CORE

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

Bell, F.R.

1963-02-01

A nuclear reactor core composed of a number of identical elements of solid moderator material fitted together was designed. Each moderator element is apertured to provide channels for fuel and coolant. The elements have an external shape which permits them to be stacked in layers with similar elements, with the surfaces of adjacent elements fitting and in contact with each other. The cross section of the element is of a general hexagonal shape with identations and protrusions, so that the elements can be fitted together. The described core should not be liable to fracture under transverse loading. Specific arrangements ofmore » moderator elements and fuel and coolant apertures are described. (M.P.G.)« less

Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

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

Spencer, Benjamin W.; Jiang, Wen; Dolbow, John E.

As a brittle material, the ceramic UO2 used as light water reactor fuel experiences significant fracturing throughout its life, beginning with the first rise to power of fresh fuel. This has multiple effects on the thermal and mechanical response of the fuel/cladding system. One such effect that is particularly important is that when there is mechanical contact between the fuel and cladding, cracks that extending from the outer surface of the fuel into the volume of the fuel cause elevated stresses in the adjacent cladding, which can potentially lead to cladding failure. Modeling the thermal and mechanical response of themore » cladding in the vicinity of these surface-breaking cracks in the fuel can provide important insights into this behavior to help avoid operating conditions that could lead to cladding failure. Such modeling has traditionally been done in the context of finite-element-based fuel performance analysis by modifying the fuel mesh to introduce discrete cracks. While this approach is effective in capturing the important behavior at the fuel/cladding interface, there are multiple drawbacks to explicitly incorporating the cracks in the finite element mesh. Because the cracks are incorporated in the original mesh, the mesh must be modified for cracks of specified location and depth, so it is difficult to account for crack propagation and the formation of new cracks at other locations. The extended finite element method (XFEM) has emerged in recent years as a powerful method to represent arbitrary, evolving, discrete discontinuities within the context of the finite element method. Development work is underway by the authors to implement XFEM in the BISON fuel performance code, and this capability has previously been demonstrated in simulations of fracture propagation in ceramic nuclear fuel. These preliminary demonstrations have included only the fuel, and excluded the cladding for simplicity. This paper presents initial results of efforts to apply XFEM to model stress concentrations induced by fuel fractures at the fuel/cladding interface during pellet cladding mechanical interaction (PCMI). This is accomplished by enhancing the thermal and mechanical contact enforcement algorithms employed by BISON to permit their use in conjunction with XFEM. The results from this methodology are demonstrated to be equivalent to those from using meshed discrete cracks. While the results of the two methods are equivalent for the case of a stationary crack, it is demonstrated that XFEM provides the additional flexibility of allowing arbitrary crack initiation and propagation during the analysis, and minimizes model setup effort for cases with stationary cracks.« less

Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature.

PubMed

Saqib, Naeem; Bäckström, Mattias

2014-12-01

Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of fly ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and chlorine content have significant effects on partitioning characteristics by increasing the formation and vaporization of highly volatile metal chlorides. Zinc and cadmium concentrations in fly ash increase with the incineration temperature. Copyright © 2014 Elsevier Ltd. All rights reserved.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-01

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

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes these overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes thses overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

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

Rowsell, David Leon

This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

Demonstration of Subscale Cermet Fuel Specimen Fabrication Approach Using Spark Plasma Sintering and Diffusion Bonding

NASA Technical Reports Server (NTRS)

Barnes, Marvin W.; Tucker, Dennis S.; Benensky, Kelsa M.

2018-01-01

Nuclear thermal propulsion (NTP) has the potential to expand the limits of human space exploration by enabling crewed missions to Mars and beyond. The viability of NTP hinges on the development of a robust nuclear fuel material that can perform in the harsh operating environment (> or = 2500K, reactive hydrogen) of a nuclear thermal rocket (NTR) engine. Efforts are ongoing to develop fuel material and to assemble fuel elements that will be stable during the service life of an NTR. Ceramic-metal (cermet) fuels are being actively pursued by NASA Marshall Space Flight Center (MSFC) due to their demonstrated high-temperature stability and hydrogen compatibility. Building on past cermet fuel development research, experiments were conducted to investigate a modern fabrication approach for cermet fuel elements. The experiments used consolidated tungsten (W)-60vol%zirconia (ZrO2) compacts that were formed via spark plasma sintering (SPS). The consolidated compacts were stacked and diffusion bonded to assess the integrity of the bond lines and internal cooling channel cladding. The assessment included hot hydrogen testing of the manufactured surrogate fuel and pure W for 45 minutes at 2500 K in the compact fuel element environmental test (CFEET) system. Performance of bonded W-ZrO2 rods was compared to bonded pure W rods to access bond line integrity and composite stability. Bonded surrogate fuels retained structural integrity throughout testing and incurred minimal mass loss.

View from southwest to northeast of fuel oil pump station, ...

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

View from southwest to northeast of fuel oil pump station, showing cooling towers to right. The tops of liquid nitrogen storage tanks A & B can be seen above the station roof. In the foreground, left to right, can be seen the covers for diesel fuel tanks no's 9 (structure #819), 8 (#818), 7 (#817), and 6 (#816). At right of center, next to the station, are no's 1 (#803) and 2 (#804). In the distant background are no's 3 (#806), 4 (#807), 5 (#808). No's 3 and 4 are 12,000-gallon tanks, the rest hold 50,000 gallons each - Stanley R. Mickelsen Safeguard Complex, Fuel Oil Pump Station, In Limited Access Area between Service Roads A & D, Nekoma, Cavalier County, ND

The manufacture of LEU fuel elements at Dounreay

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

Gibson, J.

1997-08-01

Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

Corrosion protected, multi-layer fuel cell interface

DOEpatents

Feigenbaum, Haim; Pudick, Sheldon; Wang, Chiu L.

1986-01-01

An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. The multi-layer configuration for the interface comprises a non-cupreous metal-coated metallic element to which is film-bonded a conductive layer by hot pressing a resin therebetween. The multi-layer arrangement provides bridging electrical contact.

AST Critical Propulsion and Noise Reduction Technologies for Future Commercial Subsonic Engines Area of Interest 1.0: Reliable and Affordable Control Systems

NASA Technical Reports Server (NTRS)

Myers, William; Winter, Steve

2006-01-01

The General Electric Reliable and Affordable Controls effort under the NASA Advanced Subsonic Technology (AST) Program has designed, fabricated, and tested advanced controls hardware and software to reduce emissions and improve engine safety and reliability. The original effort consisted of four elements: 1) a Hydraulic Multiplexer; 2) Active Combustor Control; 3) a Variable Displacement Vane Pump (VDVP); and 4) Intelligent Engine Control. The VDVP and Intelligent Engine Control elements were cancelled due to funding constraints and are reported here only to the state they progressed. The Hydraulic Multiplexing element developed and tested a prototype which improves reliability by combining the functionality of up to 16 solenoids and servo-valves into one component with a single electrically powered force motor. The Active Combustor Control element developed intelligent staging and control strategies for low emission combustors. This included development and tests of a Controlled Pressure Fuel Nozzle for fuel sequencing, a Fuel Multiplexer for individual fuel cup metering, and model-based control logic. Both the Hydraulic Multiplexer and Controlled Pressure Fuel Nozzle system were cleared for engine test. The Fuel Multiplexer was cleared for combustor rig test which must be followed by an engine test to achieve full maturation.

Efficient removal of nitrobenzene and concomitant electricity production by single-chamber microbial fuel cells with activated carbon air-cathode.

PubMed

Zhang, Enren; Wang, Feng; Zhai, Wenjing; Scott, Keith; Wang, Xu; Diao, Guowang

2017-04-01

Single-chamber microbial fuel cells (S-MFCs) with bio-anodes and activated carbon (AC) air-cathodes showed high nitrobenzene (NB) tolerance and NB removal with concomitant electricity production. The maximum power over 25Wm -3 could be obtained when S-MFCs were operated in the NB loading range of 1.2-6.2molm -3 d -1 , and stable electricity production over 13.7Wm -3 could be produced in a NB loading range of 1.2-14.7molm -3 d -1 . The present S-MFCs exhibited high NB removal performance with NB removal efficiency over 97% even when the NB loading rate was increased to 17.2molm -3 d -1 . The potential NB reduced product (i.e. aniline) could also be effectively removed from influents. The findings in this study means that single-chamber MFCs assembled with pre-enriched bio-anodes and AC air-cathodes could be developed as effective bio-electrochemical systems to remove NB from wastewaters and to harvest energy instead of consuming energy. Copyright © 2017 Elsevier Ltd. All rights reserved.

Comparison of the PLTEMP code flow instability predictions with measurements made with electrically heated channels for the advanced test reactor.

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

Feldman, E.

When the University of Missouri Research Reactor (MURR) was designed in the 1960s the potential for fuel element burnout by a phenomenon referred to at that time as 'autocatalytic vapor binding' was of serious concern. This type of burnout was observed to occur at power levels considerably lower than those that were known to cause critical heat flux. The conversion of the MURR from HEU fuel to LEU fuel will probably require significant design changes, such as changes in coolant channel thicknesses, that could affect the thermal-hydraulic behavior of the reactor core. Therefore, the redesign of the MURR to accommodatemore » an LEU core must address the same issues of fuel element burnout that were of concern in the 1960s. The Advanced Test Reactor (ATR) was designed at about the same time as the MURR and had similar concerns with regard to fuel element burnout. These concerns were addressed in the ATR by two groups of thermal-hydraulic tests that employed electrically heated simulated fuel channels. The Croft (1964), Reference 1, tests were performed at ANL. The Waters (1966), Reference 2, tests were performed at Hanford Laboratories in Richland Washington. Since fuel element surface temperatures rise rapidly as burnout conditions are approached, channel surface temperatures were carefully monitored in these experiments. For self-protection, the experimental facilities were designed to cut off the electric power when rapidly increasing surface temperatures were detected. In both the ATR reactor and in the tests with electrically heated channels, the heated length of the fuel plate was 48 inches, which is about twice that of the MURR. Whittle and Forgan (1967) independently conducted tests with electrically heated rectangular channels that were similar to the tests by Croft and by Walters. In the Whittle and Forgan tests the heated length of the channel varied among the tests and was between 16 and 24 inches. Both Waters and Whittle and Forgan show that the cause of the fuel element burnout is due to a form of flow instability. Whittle and Forgan provide a formula that predicts when this flow instability will occur. This formula is included in the PLTEMP/ANL code.Error! Reference source not found. Olson has shown that the PLTEMP/ANL code accurately predicts the powers at which flow instability occurs in the Whittle and Forgan experiments. He also considered the electrically heated tests performed in the ANS Thermal-Hydraulic Test Loop at ORNL and report by M. Siman-Tov et al. The purpose of this memorandum is to demonstrate that the PLTEMP/ANL code accurately predicts the Croft and the Waters tests. This demonstration should provide sufficient confidence that the PLTEMP/ANL code can adequately predict the onset of flow instability for the converted MURR. The MURR core uses light water as a coolant, has a 24-inch active fuel length, downward flow in the core, and an average core velocity of about 7 m/s. The inlet temperature is about 50 C and the peak outlet is about 20 C higher than the inlet for reactor operation at 10 MW. The core pressures range from about 4 to about 5 bar. The peak heat flux is about 110 W/cm{sup 2}. Section 2 describes the mechanism that causes flow instability. Section 3 describes the Whittle and Forgan formula for flow instability. Section 4 briefly describes both the Croft and the Waters experiments. Section 5 describes the PLTEMP/ANL models. Section 6 compares the PLTEMP/ANL predictions based on the Whittle and Forgan formula with the Croft measurements. Section 7 does the same for the Waters measurements. Section 8 provides the range of parameters for the Whittle and Forgan tests. Section 9 discusses the results and provides conclusions. In conclusion, although there is no single test that by itself closely matches the limiting conditions in the MURR, the preponderance of measured data and the ability of the Whittle and Forgan correlation, as implemented in PLTEMP/ANL, to predict the onset of flow instability for these tests leads one to the conclusion that the same method should be able to predict the onset of flow instability in the MURR reasonably well.« less

Semi-elemental formula or polymeric formula: is there a better choice for enteral nutrition in acute pancreatitis? Randomized comparative study.

PubMed

Tiengou, Laurent-Eric; Gloro, Romain; Pouzoulet, Julien; Bouhier, Karine; Read, Marie-Hélène; Arnaud-Battandier, Franck; Plaze, Jean-Marie; Blaizot, Xavier; Dao, Thông; Piquet, Marie-Astrid

2006-01-01

Jejunal nutrition is recommended during acute pancreatitis. The use of semi-elemental formulas presents several theoretical advantages over polymeric formulas, but their clinical value has been poorly documented. Our aim was to evaluate in patients with acute pancreatitis the effect of enteral nutrition by a semi-elemental formula compared with a polymeric formula. A randomized prospective pilot study, stratified according to severity, was performed in 30 consecutive patients with acute pancreatitis requiring jejunal nutrition. The semi-elemental group received 35 kcal/kg/d of Peptamen (n = 15), and the polymeric group received the same quantity of Sondalis-Iso (n = 15). Tolerance was evaluated after 7 days of enteral nutrition (D7) on visual analog scale (VAS), stool frequency, and 24-hour steatorrhea/creatorrhea. Outcome was evaluated by weight loss, length of hospital stay, and infection rate. Results were calculated as mean +/- SEM, t-test, or chi2. Patients of the 2 groups were comparable in terms of age, gender, and severity. Tolerance was good in both groups (semi-elemental vs polymeric: VAS, 7.4 +/- 0.6 vs 7.1 +/- 0.6, not significant (NS); number of stools per 24 hours, 1.7 +/- 0.4 vs 1.8 +/- 0.4, NS). Steatorrhea and creatorrhea were lower than normal in both groups. In semi-elemental group, the length of hospital stay was shorter (23 +/- 2 vs 27 +/- 1, p = .006) and weight loss was less marked (1 +/- 1 vs 2 +/- 0, p = .01). One patient in semi-elemental group and 3 patients in polymeric group developed an infection (NS). Semi-elemental and polymeric nutrition are very well tolerated in patients with acute pancreatitis. Nutrition with a semi-elemental formula supports the hypothesis of a more favorable clinical course than nutrition with a polymeric formula, but this conclusion needs to be established in larger adequately powered clinical trials.

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

Welding of unique and advanced alloys for space and high-temperature applications: welding and weldability of iridium and platinum alloys

DOE PAGES

David, Stan A.; Miller, Roger G.; Feng, Zhili

2016-08-31

Advances have been made in developing alloys for space power systems for spacecraft that travel long distances to various planets. The spacecraft are powered by radioisotope thermoelectric generators (RTGs) and the fuel element in RTGs is plutonia. For safety and containment of the radioactive fuel element, the heat source is encapsulated in iridium or platinum alloys. Ir and Pt alloys are the alloys of choice for encapsulating radioisotope fuel pellets. Ir and Pt alloys were chosen because of their high-temperature properties and compatibility with the oxide fuel element and the graphite impact shells. This review addresses the alloy design andmore » welding and weldability of Ir and Pt alloys for use in RTGs.« less

Welding of unique and advanced alloys for space and high-temperature applications: welding and weldability of iridium and platinum alloys

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

David, Stan A.; Miller, Roger G.; Feng, Zhili

Advances have been made in developing alloys for space power systems for spacecraft that travel long distances to various planets. The spacecraft are powered by radioisotope thermoelectric generators (RTGs) and the fuel element in RTGs is plutonia. For safety and containment of the radioactive fuel element, the heat source is encapsulated in iridium or platinum alloys. Ir and Pt alloys are the alloys of choice for encapsulating radioisotope fuel pellets. Ir and Pt alloys were chosen because of their high-temperature properties and compatibility with the oxide fuel element and the graphite impact shells. This review addresses the alloy design andmore » welding and weldability of Ir and Pt alloys for use in RTGs.« less

Sulphur isotope fractionation during the reduction of elemental sulphur and thiosulphate by Dethiosulfovibrio spp.

PubMed

Surkov, Alexander V; Böttcher, Michael E; Kuever, Jan

2012-01-01

Stable sulphur isotope fractionation was investigated during reduction of thiosulphate and elemental sulphur at 28°C by growing batch cultures of the sulphur- and thiosulphate-reducing bacteria Dethiosulfovibrio marinus (type strain DSM 12537) and Dethiosulfovibrio russensis (type strain DSM 12538), using citrate as carbon and energy source. The cell-specific thiosulphate reduction rate in the growth phase was 7.4±3.9 fmol cell(-1) d(-1). The hydrogen sulphide produced was enriched in (32)S by 10.3±1 ‰ compared with total thiosulphate sulphur, close to previous experimental results observed for other sulphate- and non-sulphate-reducing bacteria. Elemental sulphur reduction yields sulphur isotope enrichment factors between-1.3 and-5.2 ‰ for D. russensis and-1.7 and-5.1 ‰ for D. marinus. The smaller fractionation effects are observed in the exponential growth phase (cellular rates between 5 and 70 fmol S° cell(-1) d(-1)) and enhanced discrimination under conditions of citrate depletion and cell lysis (cellular rates between 0.3 and 3 fmol S° cell(-1) d(-1)).

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pocoima, CA; Benander, Robert E [Pacoima, CA

2010-02-23

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

HEAVY WATER MODERATED NEUTRONIC REACTOR

DOEpatents

Szilard, L.

1958-04-29

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

Compact power reactor

DOEpatents

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

1978-01-01

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

Determination of impurities in uranium matrices by time-of-flight ICP-MS using matrix-matched method

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

Buerger, Stefan; Riciputi, Lee R; Bostick, Debra A

2007-01-01

The analysis of impurities in uranium matrices is performed in a variety of fields, e.g. for quality control in the production stream converting uranium ores to fuels, as element signatures in nuclear forensics and safeguards, and for non-proliferation control. We have investigated the capabilities of time-of-flight ICP-MS for the analysis of impurities in uranium matrices using a matrix-matched method. The method was applied to the New Brunswick Laboratory CRM 124(1-7) series. For the seven certified reference materials, an overall precision and accuracy of approximately 5% and 14%, respectively, were obtained for 18 analyzed elements.

Experimental Test in a Tokamak of Fusion with Spin-Polarized D and 3He

NASA Astrophysics Data System (ADS)

Honig, Arnold; Sandorfi, Andrew

2007-06-01

An experiment to test polarization retention of highly polarized D and 3He fusion fuels prior to their fusion reactions in a tTokamak is in preparation. The fusion reaction rate with 100% vector polarized reactants is expected from simple theory to increase by a factor of 1.5. With presently available polarizations, fusion reaction enhancements of ˜15% are achievable and of significant interest, while several avenues for obtaining higher polarizations are open. The potential for survival of initial fusion fuel polarizations at ˜108 K plasma core temperatures (˜5KeV) throughout the time interval preceding fusion burn was addressed in a seminal paper in 1982. While the positive conclusion from those calculations suggests that reaction enhancements are indeed feasible, this crucial factor has never been tested in a high temperature plasma core because of difficulties in preparation and injection of sufficiently polarized fusion fuels into a high temperature reactorfusion plasma. Our solution to these problems employs a new source of highly polarized D in the form of solid HD which has been developed and used in our laboratories. Solid HD is compatible with fusion physics in view of its simplicity of elemental composition and very long (weeks) relaxation times at 4K temperature, allowing efficient polarization-preserving cold-transfer operations. Containment and polarization of the HD within polymer capsules, similar to those used in inertial confinement fusion (ICF), is an innovation which simplifies the cold-transfer of polarized fuel from the dilution refrigerator polarization-production apparatus to other liquid helium temperature cryostats, for storage, transport and placement into the barrel of a cryogenic pellet gun for firing at high velocity into the reactor. The other polarized fuel partner, 3He, has been prepared as a polarized gas for applications including high-energy polarized targets and magnetic resonance imaging (MRI) scans. It will be introduced into the reactor by loading at high pressure into a thick-walled ICF-type polymer shell for injection into the plasma core with a room temperature injection gun. Based on current experience, polarizations of both D and 3He of ˜55% are projected, producing a fusion yield increase of about 15%. A collaboration is being developed for implementing this experiment at the DIII-D Ttokamak experiment at San Diego, operated by General Atomics for the U.S. Department of Energy. Calculations indicate a 10% fusion yield increase in the 14.6 MeV protons from the D-3He reaction will provide a statistically significant test of polarization retention in the plasma. Injection of the polarized fuels into a 4He or 1H plasma improves the discrimination of the effects of polarized fuels. Details of the HD fuel preparation, of the polarization processes, and of the injection into the plasma will beare presented. If the expected fusion reaction yield increase indicative of polarization retention is detected, a route to significantly improved second generation D-3He fusion would be established, as well as confidence to undertake the more difficult polarization of tritium, which would offer important cost savings and improved prospects of ignition in the ITER program.

Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

NASA Technical Reports Server (NTRS)

Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

1974-01-01

The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

LCRE and SNAP 50-DR-1 programs. Engineering progress report, January 1, 1963--March 31, 1963

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

None

Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, primary coolant circuit, aaxiliary systems, fuel elements, instrumentation, materials development, and fabrication; and SNAP-50DR-1 specifications, fuel elements, pumps, steam generator, and materials development. (DCC)

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT ...

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

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT FRAME. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-4. INEL INDEX CODE NUMBER: 075 0701 60 851 151978. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

High temperature ceramic composition for hydrogen retention

DOEpatents

Webb, R.W.

1974-01-01

A ceramic coating for H retention in fuel elements is described. The coating has relatively low thermal neutron cross section, is not readily reduced by H at 1500 deg F, is adherent to the fuel element base metal, and is stable at reactor operating temperatures. (JRD)

JACKETED URANIUM SLUG

DOEpatents

Ohlinger, L.A.; Cooper, C.M.

1958-10-01

Fuel elements for nuclear reactors are described. Eacb fuel element is comprised of a solid cylindrical slug containing fissionable material enclosed within a fluid tight jacket of neutron permeable material such as aluminum. The jacket is provided with a flexible end cap and with a sealing member having a substantially fluid-tight fit within the jacket in tight abutment with the end cap and the end of the slug. A fluid passage is provided between the end of the slug and the cap whereby leakage fiuid is principally directed to the end of the slug. In this manner, any reaction between the fissionable material and fiuid which may take place occurs more rapidly at the end of the slug than along the sides between the slug and the jacket, thereby causing longitudinal expansion of the fuel element prior to radial expansion. The longitudinal expansion can be readily detected and the fuel element removed from the coolant tube before radial expansion causes it to become jammed in the tube.

Affordable Development and Demonstration of a Small NTR Engine and Stage: A Preliminary NASA, DOE, and Industry Assessment

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; Sefcik, Robert J.; Fittje, James E.; McCurdy, David R.; Qualls, Arthur L.; Schnitzler, Bruce G.; Werner, James E.; Weitzberg, Abraham; Joyner, Claude R.

2015-01-01

The Nuclear Thermal Rocket (NTR) represents the next evolutionary step in cryogenic liquid rocket engines. Deriving its energy from fission of uranium-235 atoms contained within fuel elements that comprise the engine's reactor core, the NTR can generate high thrust at a specific impulse of approx. 900 seconds or more - twice that of today's best chemical rockets. In FY'11, as part of the AISP project, NASA proposed a Nuclear Thermal Propulsion (NTP) effort that envisioned two key activities - "Foundational Technology Development" followed by system-level "Technology Demonstrations". Five near-term NTP activities identified for Foundational Technology Development became the basis for the NCPS project started in FY'12 and funded by NASA's AES program. During Phase 1 (FY'12-14), the NCPS project was focused on (1) Recapturing fuel processing techniques and fabricating partial length "heritage" fuel elements for the two candidate fuel forms identified by NASA and the DOE - NERVA graphite "composite" and the uranium dioxide (UO2) in tungsten "cermet". The Phase 1 effort also included: (2) Engine Conceptual Design; (3) Mission Analysis and Requirements Definition; (4) Identification of Affordable Options for Ground Testing; and (5) Formulation of an Affordable and Sustainable NTP Development Strategy. During FY'14, a preliminary plan for DDT&E was outlined by GRC, the DOE and industry for NASA HQ that involved significant system-level demonstration projects that included GTD tests at the NNSS, followed by a FTD mission. To reduce development costs, the GTD and FTD tests use a small, low thrust (approx. 7.5 or 16.5 klbf) engine. Both engines use graphite composite fuel and a "common" fuel element design that is scalable to higher thrust (approx. 25 klbf) engines by increasing the number of elements in a larger diameter core that can produce greater thermal power output. To keep the FTD mission cost down, a simple "1-burn" lunar flyby mission was considered along with maximizing the use of existing and flight proven liquid rocket and stage hardware (e.g., from the RL10-B2 engine and Delta Cryogenic Second Stage) to further ensure affordability. This paper provides a preliminary NASA, DOE and industry assessment of what is required - the key DDT&E activities, development options, and the associated schedule - to affordably build, ground test and fly a small NTR engine and stage within a 10-year timeframe.

Chemical Dissolution of Simulant FCA Cladding and Plates

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

Daniel, G.; Pierce, R.; O'Rourke, P.

The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO 3-KF) flowsheets ofmore » H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.« less

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

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

McDeavitt, Sean M

2011-04-29

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºCmore » to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow, entitled “Uranium Metal Powder Production, Particle Distribution Analysis, and Reaction Rate Studies of a Hydride-Dehydride Process"« less

78 FR 50043 - 36(b)(1) Arms Sales Notification

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-16

...-Processor (RMP) Block I 92H Missiles, 13 AN/MPQ-64F1 SENTINEL Radars, 7 AN/YSQ-184D Forward Area Air Defense... logistics support services, and other related elements of logistics support. (iv) Military Department: Army... SENTINEL Radars, 7 AN/YSQ-184D Forward Area Air Defense Command, Control, and Intelligence (FAAD C2I...

Design and characterization of a microbial fuel cell for the conversion of a lignocellulosic crop residue to electricity.

PubMed

Gregoire, K P; Becker, J G

2012-09-01

Agricultural crop residues contain high amounts of biochemical energy as cellulose and lignin. A portion of this biomass could be sustainably harvested for conversion to bioenergy to help offset fossil fuel consumption. In this study, the potential for converting lignocellulosic biomass directly to electricity in a microbial fuel cell (MFC) was explored. Design elements of tubular air cathode MFCs and leach-bed bioreactors were integrated to develop a new solid-substrate MFC in which cellulose hydrolysis, fermentation, and anode respiration occurred in a single chamber. Electricity was produced continuously from untreated corncob pellets for >60 d. Addition of rumen fluid increased power production, presumably by providing growth factors to anode-respiring bacteria. Periodic exposure to oxygen also increased power production, presumably by limiting the diversion of electrons to methanogenesis. In the absence of methanogenesis, bioaugmentation with Geobacter metallireducens further improved MFC performance. Under these conditions, the maximum power density was 230 mW/m(3). Copyright © 2012 Elsevier Ltd. All rights reserved.

Nanophase Nickel-Zirconium Alloys for Fuel Cells

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram; Whitacre, jay; Valdez, Thomas

2008-01-01

Nanophase nickel-zirconium alloys have been investigated for use as electrically conductive coatings and catalyst supports in fuel cells. Heretofore, noble metals have been used because they resist corrosion in the harsh, acidic fuel cell interior environments. However, the high cost of noble metals has prompted a search for less-costly substitutes. Nickel-zirconium alloys belong to a class of base metal alloys formed from transition elements of widely different d-electron configurations. These alloys generally exhibit unique physical, chemical, and metallurgical properties that can include corrosion resistance. Inasmuch as corrosion is accelerated by free-energy differences between bulk material and grain boundaries, it was conjectured that amorphous (glassy) and nanophase forms of these alloys could offer the desired corrosion resistance. For experiments to test the conjecture, thin alloy films containing various proportions of nickel and zirconium were deposited by magnetron and radiofrequency co-sputtering of nickel and zirconium. The results of x-ray diffraction studies of the deposited films suggested that the films had a nanophase and nearly amorphous character.

Thermal modeling in an engine cooling system to control coolant flow for fuel consumption improvement

NASA Astrophysics Data System (ADS)

Park, Sangki; Woo, Seungchul; Kim, Minho; Lee, Kihyung

2017-04-01

The design and evaluation of engine cooling and lubrication systems is generally based on real vehicle tests. Our goal here was to establish an engine heat balance model based on mathematical and interpretive analysis of each element of a passenger diesel engine cooling system using a 1-D numerical model. The purpose of this model is to determine ways of optimizing the cooling and lubrication components of an engine and then to apply these methods to actual cooling and lubrication systems of engines that will be developed in the future. Our model was operated under the New European Driving Cycle (NEDC) mode conditions, which represent the fuel economy evaluation mode in Europe. The flow rate of the cooling system was controlled using a control valve. Our results showed that the fuel efficiency was improved by as much as 1.23 %, cooling loss by 1.35 %, and friction loss by 2.21 % throughout NEDC modes by modification of control conditions.

Source Apportionment of Elemental Carbon in Beijing, China: Insights from Radiocarbon and Organic Marker Measurements.

PubMed

Zhang, Yan-Lin; Schnelle-Kreis, Jürgen; Abbaszade, Gülcin; Zimmermann, Ralf; Zotter, Peter; Shen, Rong-rong; Schäfer, Klaus; Shao, Longyi; Prévôt, André S H; Szidat, Sönke

2015-07-21

Elemental carbon (EC) or black carbon (BC) in the atmosphere has a strong influence on both climate and human health. In this study, radiocarbon ((14)C) based source apportionment is used to distinguish between fossil fuel and biomass burning sources of EC isolated from aerosol filter samples collected in Beijing from June 2010 to May 2011. The (14)C results demonstrate that EC is consistently dominated by fossil-fuel combustion throughout the whole year with a mean contribution of 79% ± 6% (ranging from 70% to 91%), though EC has a higher mean and peak concentrations in the cold season. The seasonal molecular pattern of hopanes (i.e., a class of organic markers mainly emitted during the combustion of different fossil fuels) indicates that traffic-related emissions are the most important fossil source in the warm period and coal combustion emissions are significantly increased in the cold season. By combining (14)C based source apportionment results and picene (i.e., an organic marker for coal emissions) concentrations, relative contributions from coal (mainly from residential bituminous coal) and vehicle to EC in the cold period were estimated as 25 ± 4% and 50 ± 7%, respectively, whereas the coal combustion contribution was negligible or very small in the warm period.

Nuclear fuel element

DOEpatents

Meadowcroft, Ronald Ross; Bain, Alastair Stewart

1977-01-01

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

Fuel cell generator energy dissipator

DOEpatents

Veyo, Stephen Emery; Dederer, Jeffrey Todd; Gordon, John Thomas; Shockling, Larry Anthony

2000-01-01

An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a fuel cell generator when the electrical power output of the fuel cell generator is terminated. During a generator shut down condition, electrically resistive elements are automatically connected across the fuel cell generator terminals in order to draw current, thereby depleting the fuel

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

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

1993-09-01

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

Pyrochlore-type catalysts for the reforming of hydrocarbon fuels

DOEpatents

Berry, David A [Morgantown, WV; Shekhawat, Dushyant [Morgantown, WV; Haynes, Daniel [Morgantown, WV; Smith, Mark [Morgantown, WV; Spivey, James J [Baton Rouge, LA

2012-03-13

A method of catalytically reforming a reactant gas mixture using a pyrochlore catalyst material comprised of one or more pyrochlores having the composition A.sub.2-w-xA'.sub.wA''.sub.xB.sub.2-y-zB'.sub.yB''.sub.zO.sub.7-.DELTA.. Distribution of catalytically active metals throughout the structure at the B site creates an active and well dispersed metal locked into place in the crystal structure. This greatly reduces the metal sintering that typically occurs on supported catalysts used in reforming reactions, and reduces deactivation by sulfur and carbon. Further, oxygen mobility may also be enhanced by elemental exchange of promoters at sites in the pyrochlore. The pyrochlore catalyst material may be utilized in catalytic reforming reactions for the conversion of hydrocarbon fuels into synthesis gas (H.sub.2+CO) for fuel cells, among other uses.

Pyrochlore catalysts for hydrocarbon fuel reforming

DOEpatents

Berry, David A.; Shekhawat, Dushyant; Haynes, Daniel; Smith, Mark; Spivey, James J.

2012-08-14

A method of catalytically reforming a reactant gas mixture using a pyrochlore catalyst material comprised of one or more pyrochlores having the composition A₂B_2-y-zB'_yB"_zO_7-.DELTA., where y>0 and z.gtoreq.0. Distribution of catalytically active metals throughout the structure at the B site creates an active and well dispersed metal locked into place in the crystal structure. This greatly reduces the metal sintering that typically occurs on supported catalysts used in reforming reactions, and reduces deactivation by sulfur and carbon. Further, oxygen mobility may also be enhanced by elemental exchange of promoters at sites in the pyrochlore. The pyrochlore catalyst material may be utilized in catalytic reforming reactions for the conversion of hydrocarbon fuels into synthesis gas (H₂+CO) for fuel cells, among other uses.

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

Studies of behavior of the fuel compound based on the U-Zr micro-heterogeneous quasialloy during cyclic thermal tests

NASA Astrophysics Data System (ADS)

Zaytsev, D. A.; Repnikov, V. M.; Soldatkin, D. M.; Solntsev, V. A.

2017-11-01

This paper provides the description of temperature cycle testing of U-Zr heterogeneous fuel composition. The composition is essentially a niobium-doped zirconium matrix with metallic uranium filaments evenly distributed over the cross section. The test samples 150 mm long had been fabricated using a fiber-filament technology. The samples were essentially two-bladed spiral mandrel fuel elements parts. In the course of experiments the following temperatures were applied: 350, 675, 780 and 1140 °C with total exposure periods equal to 200, 30, 30 and 6 hours respectively. The fuel element samples underwent post-exposure material science examination including: geometry measurements, metallographic analysis, X-ray phase analysis and electron-microscopic analysis as well as micro-hardness measurement. It has been found that no significant thermal swelling of the samples occurs throughout the whole temperature range from 350 °C up to 1140 °C. The paper presents the structural changes and redistribution of the fuel component over the fuel element cross section with rising temperature.

Restartable High Power Gas Generator.

DTIC Science & Technology

1982-12-01

RELEASE; DISTRIBUTION UNLIMITED DTIC AERO PROPULSION LABORATORY " D T-C ) AIR FORCE WRIGHT AERONAUTICAL LABORATORIES AIR FORCE SYSTEMS COMMAND APR 2 3 133 C...INIFRV 4umat OVT CCESIONNO. 3 .PERCORMNG’ CAORGO NUMER 14. ATLE(ad S.##* CTPOfRCTOR GR N OERED M. G. Gants F33615-79-C-2004 9. PERFORMING ORGANIZATION...Assembly 11 3 Injector Internal Confiquration 12 4 Injector Assembly 15 5 Injector Housing le 6 Pintle 17 7 Core 18 8 Fuel Injection Rina iS 9 Fuel

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE PAGES

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael; ...

2016-11-17

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

DOEpatents

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

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

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

SHAPED FISSIONABLE METAL BODIES

DOEpatents

Wigner, E.P.; Williamson, R.R.; Young, G.J.

1958-10-14

A technique is presented for grooving the surface of fissionable fuel elements so that expansion can take place without damage to the interior structure of the fuel element. The fissionable body tends to develop internal stressing when it is heated internally by the operation of the nuclear reactor and at the same time is subjected to surface cooling by the circulating coolant. By producing a grooved or waffle-like surface texture, the annular lines of tension stress are disrupted at equally spaced intervals by the grooves, thereby relieving the tension stresses in the outer portions of the body while also facilitating the removal of accumulated heat from the interior portion of the fuel element.

Study of Laminar Flame 2-D Scalar Values at Various Fuel to Air Ratios Using an Imaging Fourier-Transform Spectrometer and 2-D CFD Analysis

DTIC Science & Technology

2013-03-01

NASA- Glenn’s Chemical Equilibrium with Applications (CEA) program. UNICORN CFD predictions were in excellent agreement with CEA calculations at...49 Appendix A – UNICORN CFD Inputs and Instruction .....................................................50 Appendix B – NASA-Glenn...17 Figure 7: Schematic of UNICORN CFD card setup. ........................................................ 18 Figure 8: Averaged flame

FUEL ELEMENT INTERLOCKING ARRANGEMENT

DOEpatents

Fortescue, P.; Nicoll, D.

1963-01-01

This patent relates to a system for mutually interlocking a multiplicity of elongated, parallel, coextensive, upright reactor fuel elements so as to render a laterally selfsupporting bundle, while admitting of concurrent, selective, vertical withdrawal of a sizeable number of elements without any of the remaining elements toppling, Each element is provided with a generally rectangular end cap. When a rank of caps is aligned in square contact, each free edge centrally defines an outwardly profecting dovetail, and extremitally cooperates with its adjacent cap by defining a juxtaposed half of a dovetail- receptive mortise. Successive ranks are staggered to afford mating of their dovetails and mortises. (AEC)

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

NASA Astrophysics Data System (ADS)

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

2017-04-01

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

NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.; Kopelman, B.; Hausner, H.H.

1963-07-01

A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

Emission factors for gaseous and particulate pollutants from offshore diesel engine vessels in China

NASA Astrophysics Data System (ADS)

Zhang, Fan; Chen, Yingjun; Tian, Chongguo; Lou, Diming; Li, Jun; Zhang, Gan; Matthias, Volker

2016-05-01

Shipping emissions have significant influence on atmospheric environment as well as human health, especially in coastal areas and the harbour districts. However, the contribution of shipping emissions on the environment in China still need to be clarified especially based on measurement data, with the large number ownership of vessels and the rapid developments of ports, international trade and shipbuilding industry. Pollutants in the gaseous phase (carbon monoxide, sulfur dioxide, nitrogen oxides, total volatile organic compounds) and particle phase (particulate matter, organic carbon, elemental carbon, sulfates, nitrate, ammonia, metals) in the exhaust from three different diesel-engine-powered offshore vessels in China (350, 600 and 1600 kW) were measured in this study. Concentrations, fuel-based and power-based emission factors for various operating modes as well as the impact of engine speed on emissions were determined. Observed concentrations and emission factors for carbon monoxide, nitrogen oxides, total volatile organic compounds, and particulate matter were higher for the low-engine-power vessel (HH) than for the two higher-engine-power vessels (XYH and DFH); for instance, HH had NOx EF (emission factor) of 25.8 g kWh-1 compared to 7.14 and 6.97 g kWh-1 of DFH, and XYH, and PM EF of 2.09 g kWh-1 compared to 0.14 and 0.04 g kWh-1 of DFH, and XYH. Average emission factors for all pollutants except sulfur dioxide in the low-engine-power engineering vessel (HH) were significantly higher than that of the previous studies (such as 30.2 g kg-1 fuel of CO EF compared to 2.17 to 19.5 g kg-1 fuel in previous studies, 115 g kg-1 fuel of NOx EF compared to 22.3 to 87 g kg-1 fuel in previous studies and 9.40 g kg-1 fuel of PM EF compared to 1.2 to 7.6 g kg-1 fuel in previous studies), while for the two higher-engine-power vessels (DFH and XYH), most of the average emission factors for pollutants were comparable to the results of the previous studies, engine type was one of the most important influence factors for the differences. Emission factors for all three vessels were significantly different during different operating modes. Organic carbon and elemental carbon were the main components of particulate matter, while water-soluble ions and elements were present in trace amounts. The test inland ships and some test offshore vessels in China always had higher EFs for CO, NOx, and PM than previous studies. Besides, due to the significant influence of engine type on shipping emissions and that no accurate local EFs could be used in inventory calculation, much more measurement data for different vessels in China are still in urgent need. Best-fit engine speeds during actual operation should be based on both emission factors and economic costs.

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Hauth, J.J.; Anicetti, R.J.

1962-12-01

A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

FOIL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Noland, R.A.; Walker, D.E.; Spinrad, B.I.

1963-07-16

A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

The EPA National Fuels Surveillance Network. I. Trace constituents in gasoline and commercial gasoline fuel additives.

PubMed Central

Jungers, R H; Lee, R E; von Lehmden, D J

1975-01-01

A National Fuels Surveillance Network has been established to collect gasoline and other fuels through the 10 regional offices of the Environmental Protection Agency. Physical, chemical, and trace element analytical determinations are made on the collected fuel samples to detect components which may present an air pollution hazard or poison exhaust catalytic control devices. A summary of trace elemental constituents in over 50 gasoline samples and 18 commercially marketed consumer purchased gasoline additives is presented. Quantities of Mn, Ni, Cr, Zn, Cu, Fe, Sb, B, Mg, Pb, and S were found in most regular and premium gasoline. Environmental implications of trace constituents in gasoline are discussed. PMID:1157783

Performance assessment of 700-bar compressed hydrogen storage for light duty fuel cell vehicles

DOE PAGES

Hua, Thanh Q.; Roh, Hee-Seok; Ahluwalia, Rajesh K.

2017-09-11

In this study, type 4 700-bar compressed hydrogen storage tanks were modeled using ABAQUS. The finite element model was first calibrated against data for 35-L subscale test tanks to obtain the composite translation efficiency, and then applied to full sized tanks. Two variations of the baseline T700/epoxy composite were considered in which the epoxy was replaced with a low cost vinyl ester resin and low cost resin with an alternate sizing. The results showed that the reduction in composite weight was attributed primarily to the lower density of the resin and higher fiber volume fraction in the composite due tomore » increased squeeze-out with the lower viscosity vinyl ester resin. The system gravimetric and volumetric capacities for the onboard storage system that holds 5.6 kg H 2 are 4.2 wt% (1.40 kWh/kg) and 24.4 g-H 2/L (0.81 kWh/L), respectively. The system capacities increase and carbon fiber requirement decreases if the in-tank amount of unrecoverable hydrogen is reduced by lowering the tank "empty" pressure. Models of an alternate tank design showed potential 4-7% saving in composite usage for tanks with a length-to-diameter (L/D) ratio of 2.8-3.0 but no saving for L/D of 1.7. Lastly, a boss with smaller opening and longer flange does not appear to reduce the amount of helical windings.« less

Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

NASA Astrophysics Data System (ADS)

Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

2017-07-01

Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10-6 K-1 (α∥) and 6.15 × 10-6 K-1 (α⊥) at the temperature range of 25-700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

Gaseous and Particulate Emissions from Diesel Engines at Idle and under Load: Comparison of Biodiesel Blend and Ultralow Sulfur Diesel Fuels

PubMed Central

Chin, Jo-Yu; Batterman, Stuart A.; Northrop, William F.; Bohac, Stanislav V.; Assanis, Dennis N.

2015-01-01

Diesel exhaust emissions have been reported for a number of engine operating strategies, after-treatment technologies, and fuels. However, information is limited regarding emissions of many pollutants during idling and when biodiesel fuels are used. This study investigates regulated and unregulated emissions from both light-duty passenger car (1.7 L) and medium-duty (6.4 L) diesel engines at idle and load and compares a biodiesel blend (B20) to conventional ultralow sulfur diesel (ULSD) fuel. Exhaust aftertreatment devices included a diesel oxidation catalyst (DOC) and a diesel particle filter (DPF). For the 1.7 L engine under load without a DOC, B20 reduced brake-specific emissions of particulate matter (PM), elemental carbon (EC), nonmethane hydrocarbons (NMHCs), and most volatile organic compounds (VOCs) compared to ULSD; however, formaldehyde brake-specific emissions increased. With a DOC and high load, B20 increased brake-specific emissions of NMHC, nitrogen oxides (NOx), formaldehyde, naphthalene, and several other VOCs. For the 6.4 L engine under load, B20 reduced brake-specific emissions of PM2.5, EC, formaldehyde, and most VOCs; however, NOx brake-specific emissions increased. When idling, the effects of fuel type were different: B20 increased NMHC, PM2.5, EC, formaldehyde, benzene, and other VOC emission rates from both engines, and changes were sometimes large, e.g., PM2.5 increased by 60% for the 6.4 L/2004 calibration engine, and benzene by 40% for the 1.7 L engine with the DOC, possibly reflecting incomplete combustion and unburned fuel. Diesel exhaust emissions depended on the fuel type and engine load (idle versus loaded). The higher emissions found when using B20 are especially important given the recent attention to exposures from idling vehicles and the health significance of PM2.5. The emission profiles demonstrate the effects of fuel type, engine calibration, and emission control system, and they can be used as source profiles for apportionment, inventory, and exposure purposes. PMID:25722535

Evaluation of Systems Engineering Methods, Processes and Tools on Department of Defense and Intelligence Community Programs - Phase 2

DTIC Science & Technology

2009-12-15

O 0 0 1, T O 0 0 2...ber: H 9 8 230 -0 8 -D -0 171 D O 0 0 1, T O 0 0 2. R T 0 0 9 4 7 C O N TIN UO US IN TEG RA TIO N Theme Elements...D O 0 0 1, T O 0 0 2. R T 0 0 9 51 RA PID PRO TO TYPIN G Theme Elements Time-boxed development process Delivery

26 CFR 48.4091-3 - Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d).

Code of Federal Regulations, 2013 CFR

2013-04-01

... 26 Internal Revenue 16 2013-04-01 2013-04-01 false Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d). 48.4091-3 Section 48.4091-3 Internal Revenue INTERNAL... Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d). (a) Overview...

26 CFR 48.4091-3 - Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d).

Code of Federal Regulations, 2010 CFR

2010-04-01

... 26 Internal Revenue 16 2010-04-01 2010-04-01 true Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d). 48.4091-3 Section 48.4091-3 Internal Revenue INTERNAL... Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d). (a) Overview...

26 CFR 48.4091-3 - Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d).

Code of Federal Regulations, 2011 CFR

2011-04-01

... 26 Internal Revenue 16 2011-04-01 2011-04-01 false Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d). 48.4091-3 Section 48.4091-3 Internal Revenue INTERNAL... Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d). (a) Overview...

26 CFR 48.4091-3 - Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d).

Code of Federal Regulations, 2012 CFR

2012-04-01

... 26 Internal Revenue 16 2012-04-01 2012-04-01 false Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d). 48.4091-3 Section 48.4091-3 Internal Revenue INTERNAL... Aviation fuel; conditions to allowance of refunds of aviation fuel tax under section 4091(d). (a) Overview...

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

2011-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

2010-04-01

The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.