Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

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

Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

2016-01-01

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermalmore » hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.« less

Method and etchant to join Ag-clad BSSCO superconducting tape

DOEpatents

Balachandran, U.; Iyer, A.N.; Huang, J.Y.

1999-03-16

A method of removing a silver cladding from high temperature superconducting material clad in silver (HTS) is disclosed. The silver clad HTS is contacted with an aqueous solution of HNO{sub 3} followed by an aqueous solution of NH{sub 4}OH and H{sub 2}O{sub 2} for a time sufficient to remove the silver cladding from the superconducting material without adversely affecting the superconducting properties of the superconducting material. A portion of the silver cladding may be masked with a material chemically impervious to HNO{sub 3} and to a combination of NH{sub 4}OH and H{sub 2}O{sub 2} to preserve the Ag coating. A silver clad superconductor is disclosed, made in accordance with the method discussed. 3 figs.

Nuclear reactor fuel element

DOEpatents

Johnson, Carl E.; Crouthamel, Carl E.

1980-01-01

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

Early implementation of SiC cladding fuel performance models in BISON

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

Powers, Jeffrey J.

2015-09-18

SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation duemore » to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.« less

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.

Posttest data analysis of FIST experimental TRAC-BD1/MOD1 power transient experiment

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

Wheatley, P.D.; Wagner, K.C.

The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting in only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena: (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

Research on Microstructure and Property of TiC-Co Composite Material Made by Laser Cladding

NASA Astrophysics Data System (ADS)

Zhang, Wei

The experiment of laser cladding on the surface of 2Cr13 steel was made. Titanium carbide (TiC) powder and Co-base alloy powder were used as cladding material. The microstructure and property of laser cladding layer were tested. The research showed that laser cladding layer had better properties such as minute crystals, deeper layer, higher hardness and good metallurgical bonding with base metal. The structure of cladding was supersaturated solid solution with dispersed titanium carbide. The average hardness of cladding zone was 660HV0.2. 2Cr13 steel was widely used in the field of turbine blades. Using laser cladding, the good wear layer would greatly increase the useful life of turbine blades.

Screening of advanced cladding materials and UN-U3Si5 fuel

NASA Astrophysics Data System (ADS)

Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

2015-07-01

In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in these assessments are preliminary, and that additional data are necessary for these materials, most significantly under irradiation.

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.

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

DOE PAGES

George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; ...

2014-09-29

A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in themore » fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the harder neutron spectrum in the system, causing more 239Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Furthermore, implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235U enrichment and the additional UO 2 pellet volume enabled by using thinner cladding.« less

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

Lee, Young-Ho; Byun, Thak Sang

Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate themore » tribological compatibilities of selfmated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.« less

Nuclear reactor fuel element with vanadium getter on cladding

DOEpatents

Johnson, Carl E.; Carroll, Kenneth G.

1977-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

High burn-up spent nuclear fuel transport reliability investigation

DOE PAGES

Wang, Jy-An; Wang, Hong; Jiang, Hao; ...

2018-04-15

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which hasmore » provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. The data presented here can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.« less

High burn-up spent nuclear fuel transport reliability investigation

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

Wang, Jy-An; Wang, Hong; Jiang, Hao

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which hasmore » provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. The data presented here can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.« less

Corrosion inhibition of steam generator tubesheet by Alloy 690 cladding in secondary side environments

NASA Astrophysics Data System (ADS)

Hur, Do Haeng; Choi, Myung Sik; Lee, Deok Hyun; Han, Jung Ho; Shim, Hee Sang

2013-11-01

Denting is a phenomenon that a steam generator tube is distorted by a volume expansion of corrosion products of the tube support and tubesheet materials adjacent to the tube. Although denting has been mitigated by a modification of the design and material of the tube support structures, it has been an inevitable concern in the crevice region of the top of tubesheet. This paper provides a new technology to prevent denting by cladding the secondary surface of the tubesheet with a corrosion resistant material. In this study, Alloy 690 material was cladded onto the surface of an SA508 tubesheet to a thickness of about 9 mm. The corrosion rates of the original SA508 tubesheet and the Alloy 690 clad material were measured in acidic and alkaline simulated environments. Using Alloy 690 cladding, the corrosion rate of the tubesheet within a magnetite sludge pile decreased by a factor of 680 in 0.1 M NiCl2 solution at 300 °C, and by a factor of 58 in 2 M NaOH solution at 315 °C. This means that denting can drastically be prevented by cladding the secondary tubesheet surface with corrosion resistant materials.

Boron-copper neutron absorbing material and method of preparation

DOEpatents

Wiencek, Thomas C.; Domagala, Robert F.; Thresh, Henry

1991-01-01

A composite, copper clad neutron absorbing material is comprised of copper powder and boron powder enriched with boron 10. The boron 10 content can reach over 30 percent by volume, permitting a very high level of neutron absorption. The copper clad product is also capable of being reduced to a thickness of 0.05 to 0.06 inches and curved to a radius of 2 to 3 inches, and can resist temperatures of 900.degree. C. A method of preparing the material includes the steps of compacting a boron-copper powder mixture and placing it in a copper cladding, restraining the clad assembly in a steel frame while it is hot rolled at 900.degree. C. with cross rolling, and removing the steel frame and further rolling the clad assembly at 650.degree. C. An additional sheet of copper can be soldered onto the clad assembly so that the finished sheet can be cold formed into curved shapes.

Severe accident modeling of a PWR core with different cladding materials

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

Johnson, S. C.; Henry, R. E.; Paik, C. Y.

2012-07-01

The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCSmore » rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)« less

SiC Composite for Fuel Structure Applications

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

Yueh, Ken

Extensive evaluation was performed to determine the suitability of using SiC composite as a boiling water reactor (BWR) fuel channel material. A thin walled SiC composite box, 10 cm in dimension by approximately 1.5 mm wall thickness was fabricated using chemical vapor deposition (CVD) for testing. Mechanical test results and performance evaluations indicate the material could meet BWR channel mechanical design requirement. However, large mass loss of up to 21% was measured in in-pile corrosion test under BWR-like conditions in under 3 months of irradiation. A fresh sister sample irradiated in a follow-up cycle under PWR conditions showed no measureablemore » weight loss and thus supports the hypothesis that the oxidizing condition of the BWR-like coolant chemistry was responsible for the high corrosion rate. A thermodynamic evaluation showed SiC is not stable and the material may oxidize to form SiO 2 and CO 2. Silica has demonstrated stability in high temperature steam environment and form a protective oxide layer under severe accident conditions. However, it does not form a protective layer in water under normal BWR operational conditions due to its high solubility. Corrosion product stabilization by modifying the SiC CVD surface is an approach evaluated in this study to mitigate the high corrosion rate. Titanium and zirconium have been selected as stabilizing elements since both TiSiO 4 and ZrSiO 4 are insoluble in water. Corrosion test results in oxygenated water autoclave indicate TiSiO4 does not form a protective layer. However, zirconium doped test samples appear to form a stable continuous layer of ZrSiO 4 during the corrosion process. Additional process development is needed to produce a good ZrSiC coating to verify functionality of the mitigation concept.« less

Posttest data analysis and assessment of TRAC-BD1/MOD1 with data from a Full Integral Simulation Test (FIST) power transient experiment

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

Wheatley, P.D.; Wagner, K.C.

The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting on only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena; (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

Nuclear fuel performance: Trends, remedies and challenges

NASA Astrophysics Data System (ADS)

Rusch, C. A.

2008-12-01

It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

NASA Astrophysics Data System (ADS)

Yano, Y.; Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T.; Ukai, S.; Oono, N.; Kimura, A.; Hayashi, S.; Torimaru, T.

2017-04-01

Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900-1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

Cladding material, tube including such cladding material and methods of forming the same

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

Garnier, John E.; Griffith, George W.

A multi-layered cladding material including a ceramic matrix composite and a metallic material, and a tube formed from the cladding material. The metallic material forms an inner liner of the tube and enables hermetic sealing of thereof. The metallic material at ends of the tube may be exposed and have an increased thickness enabling end cap welding. The metallic material may, optionally, be formed to infiltrate voids in the ceramic matrix composite, the ceramic matrix composite encapsulated by the metallic material. The ceramic matrix composite includes a fiber reinforcement and provides increased mechanical strength, stiffness, thermal shock resistance and highmore » temperature load capacity to the metallic material of the inner liner. The tube may be used as a containment vessel for nuclear fuel used in a nuclear power plant or other reactor. Methods for forming the tube comprising the ceramic matrix composite and the metallic material are also disclosed.« less

Methodology for Mechanical Property Testing of Fuel Cladding Using a Expanded Plug Wedge Test

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

Jiang, Hao; Wang, Jy-An John

2014-01-01

An expanded plug method was developed earlier for determining the tensile properties of irradiated fuel cladding. This method tests fuel rod cladding ductility by utilizing an expandable plug to radially stretch a small ring of irradiated cladding material. The circumferential or hoop strain is determined from the measured diametrical expansion of the ring. A developed procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves, from which material properties of the cladding can be extracted. However, several deficiencies existed in this expanded-plug test that can impact the accuracy of test results, suchmore » as that the large axial compressive stress resulted from the expansion plug test can potentially induce the shear failure mode of the tested specimen. Moreover, highly nonuniform stress and strain distribution in the deformed clad gage section and significant compressive stresses, induced by bending deformation due to clad bulging effect, will further result in highly nonconservative estimates of the mechanical properties for both strength and ductility of the tested clad. To overcome the aforementioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. By optimizing the specific geometry designs, selecting the appropriate material for the expansion plug, and adding new components into the testing system, a modified expansion plug testing protocol has been developed. A general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor, -factor, was used to convert the ring load Fring into hoop stress , and is written as _ = F_ring/tl , where t is the clad thickness and l is the clad length. The generated stress-strain curve agrees well with the associated tensile test data in both elastic and plastic deformation regions.« less

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)

46 CFR 111.60-23 - Metal-clad (Type MC) cable.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Metal-clad (Type MC) cable. 111.60-23 Section 111.60-23...-GENERAL REQUIREMENTS Wiring Materials and Methods § 111.60-23 Metal-clad (Type MC) cable. (a) Metal-clad (Type MC) cable permitted on board a vessel must be continuous corrugated metal-clad cable. (b) The...

46 CFR 111.60-23 - Metal-clad (Type MC) cable.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 46 Shipping 4 2011-10-01 2011-10-01 false Metal-clad (Type MC) cable. 111.60-23 Section 111.60-23...-GENERAL REQUIREMENTS Wiring Materials and Methods § 111.60-23 Metal-clad (Type MC) cable. (a) Metal-clad (Type MC) cable permitted on board a vessel must be continuous corrugated metal-clad cable. (b) The...

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Kazimi, Mujid S.

2013-10-01

The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

Characterizing the effects of cladding on semi-elliptical longitudinal surface flaws in cylindrical vessels subjected to internal pressure

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

Killian, D.E.; Yoon, K.K.

1996-12-01

Flaws on the inside surface of cladded reactor vessels are often analyzed by modelling the carbon steel base metal without consideration of a layer of stainless steel cladding material, thus ignoring the effects of this bimetallic discontinuity. Adding cladding material to the inside surface of a finite element model of a vessel raises concerns regarding adequate mesh refinement in the vicinity of the base metal/cladding interface. This paper presents results of three-dimensional linear stress analysis that has been performed to obtain stress intensity factors for clad and unclad reactor vessels subjected to internal pressure loading. The study concentrates on semi-ellipticalmore » longitudinal surface flaws with a 6 to 1 length-to-depth ratio and flaw depths of 1/8 and 1/4 of the base metal thickness. Various meshing schemes are evaluated for modelling the crack front profile, with particular emphasis on the region near the inside surface and at the base metal/cladding interface. The shape of the crack front profile through the cladding layer and the number of finite elements used to discretize the cladding thickness are found to have a significant influence on typical fracture mechanic measures of the crack tip stress fields. Results suggest that the stress intensity factor at the inner surface of a cladded vessel may be affected as much by the finite element mesh near the surface as by the material discontinuity between the two parts of the structure.« less

Qualification of submerged-arc narrow strip cladding process

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

Ayres, P.S.; Gottschling, J.D.; Jeffers, G.K.

1975-08-01

An unique narrow strip cladding process for use on both plate and forging material for nuclear components was developed. The qualification testing of this low-heat input process for cladding nuclear components, including those of SA508 Class 2 material is described. The theory that explains the acceptable results of these tests is also given. (auth)

Qualification of submerged-arc narrow strip cladding process

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

Ayres, P.S.; Gottschling, J.D.; Jeffers, G.K.

1976-03-01

Babcock and Wilcox has developed an unique narrow strip cladding process for use on both plate and forging material for nuclear components. The qualification testing of this low-heat input process for cladding nuclear components is described, including those of SA508 Class 2 material. The theory that explains the acceptable results of these tests is also given.

Thermal effects on transducer material for heat assisted magnetic recording application

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

Ji, Rong, E-mail: Ji-Rong@dsi.a-star.edu.sg; Xu, Baoxi; Cen, Zhanhong

2015-05-07

Heat Assisted Magnetic Recording (HAMR) is a promising technology for next generation hard disk drives with significantly increased data recording capacities. In HAMR, an optical near-field transducer (NFT) is used to concentrate laser energy on a magnetic recording medium to fulfill the heat assist function. The key components of a NFT are transducer material, cladding material, and adhesion material between the cladding and the transducer materials. Since transducer materials and cladding materials have been widely reported, this paper focuses on the adhesion materials between the Au transducer and the Al{sub 2}O{sub 3} cladding material. A comparative study for two kindsmore » of adhesion material, Ta and Cr, has been conducted. We found that Ta provides better thermal stability to the whole transducer than Cr. This is because after thermal annealing, chromium forms oxide material at interfaces and chromium atoms diffuse remarkably into the Au layer and react with Au to form Au alloy. This study also provides insights on the selection of adhesion material for HAMR transducer.« less

Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

NASA Astrophysics Data System (ADS)

Li, Bo-Shiuan

Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission gas release. An optimum design is sought considering both thermal and mechanical models of this ceramic cladding with UO2 and advanced high density fuels.

High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

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

McHugh, Kevin M; Garnier, John E; Sergey Rashkeev

2013-01-01

Stability of cladding materials under off-normal conditions is an important issue for the safe operation of light water nuclear reactors. Metals, ceramics, and metal/ceramic composites are being investigated as substitutes for traditional zirconium-based cladding. To support down-selection of these advanced materials and designs, a test apparatus was constructed to study the onset and evolution of cladding oxidation, and deformation behavior of cladding materials, under loss-of-coolant accident scenarios. Preliminary oxidation tests were conducted in dry oxygen and in saturated steam/air environments at 1000OC. Tube samples of Zr-702, Zr-702 reinforced with 1 ply of a ß-SiC CMC overbraid, and sintered a-SiC weremore » tested. Samples were induction heated by coupling to a molybdenum susceptor inside the tubes. The deformation behavior of He-pressurized tubes of Zr-702 and SiC CMC-reinforced Zr-702, heated to rupture, was also examined.« less

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1983-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

NASA Astrophysics Data System (ADS)

Shirvan, Koroush

Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater temperature was kept the same for the BWR-HD and ABWR which resulted in 4 °K cooler core inlet temperature for the BWR-HD given that its feedwater makes up a larger fraction of total core flow. The stability analysis using the STAB and S3K codes showed satisfactory results for the hot channel, coupled regional out-of-phase and coupled core-wide in-phase modes. A RELAPS model of the ABWR system was constructed and applied to six transients for the BWR-HD and ABWR. The 6MCPRs during all the transients were found to be equal or less for the new design and the core remained covered for both. The lower void coefficient along with smaller core volume proved to be advantages for the simulated transients. Helical Cruciform Fuel (HCF) rods were proposed in prior MIT studies to enhance the fuel surface to volume ratio. In this work, higher fidelity models (e.g. CFD instead of subchannel methods for the hydraulic behaviour) are used to investigate the resolution needed for accurate assessment of the HCF design. For neutronics, conserving the fuel area of cylindrical rods results in a different reactivity level with a lower void coefficient for the HCF design. In single-phase flow, for which experimental results existed, the friction factor is found to be sensitive to HCF geometry and cannot be calculated using current empirical models. A new approach for analysis of flow crisis conditions for HCF rods in the context of Departure from Nucleate Boiling (DNB) and dryout using the two phase interface tracking method was proposed and initial results are presented. It is shown that the twist of the HCF rods promotes detachment of a vapour bubble along the elbows which indicates no possibility for an early DNB for the HCF rods and in fact a potential for a higher DNB heat flux. Under annular flow conditions, it was found that the twist suppressed the liquid film thickness on the HCF rods, at the locations of the highest heat flux, which increases the possibility of reaching early dryout. It was also shown that modeling the 3D heat and stress distribution in the HCF rods is necessary for accurate steady state and transient analyses. (Abstract shortened by UMI.) (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

Accident tolerant fuel cladding development: Promise, status, and challenges

NASA Astrophysics Data System (ADS)

Terrani, Kurt A.

2018-04-01

The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

Standard-less analysis of Zircaloy clad samples by an instrumental neutron activation method

NASA Astrophysics Data System (ADS)

Acharya, R.; Nair, A. G. C.; Reddy, A. V. R.; Goswami, A.

2004-03-01

A non-destructive method for analysis of irregular shape and size samples of Zircaloy has been developed using the recently standardized k0-based internal mono standard instrumental neutron activation analysis (INAA). The samples of Zircaloy-2 and -4 tubes, used as fuel cladding in Indian boiling water reactors (BWR) and pressurized heavy water reactors (PHWR), respectively, have been analyzed. Samples weighing in the range of a few tens of grams were irradiated in the thermal column of Apsara reactor to minimize neutron flux perturbations and high radiation dose. The method utilizes in situ relative detection efficiency using the γ-rays of selected activation products in the sample for overcoming γ-ray self-attenuation. Since the major and minor constituents (Zr, Sn, Fe, Cr and/or Ni) in these samples were amenable to NAA, the absolute concentrations of all the elements were determined using mass balance instead of using the concentration of the internal mono standard. Concentrations were also determined in a smaller size Zircaloy-4 sample by irradiating in the core position of the reactor to validate the present methodology. The results were compared with literature specifications and were found to be satisfactory. Values of sensitivities and detection limits have been evaluated for the elements analyzed.

Experimental and numerical investigation on cladding of corrosion-erosion resistant materials by a high power direct diode laser

NASA Astrophysics Data System (ADS)

Farahmand, Parisa

In oil and gas industry, soil particles, crude oil, natural gas, particle-laden liquids, and seawater can carry various highly aggressive elements, which accelerate the material degradation of component surfaces by combination of slurry erosion, corrosion, and wear mechanisms. This material degradation results into the loss of mechanical properties such as strength, ductility, and impact strength; leading to detachment, delamination, cracking, and ultimately premature failure of components. Since the failure of high valued equipment needs considerable cost and time to be repaired or replaced, minimizing the tribological failure of equipment under aggressive environment has been gaining increased interest. It is widely recognized that effective management of degradation mechanisms will contribute towards the optimization of maintenance, monitoring, and inspection costs. The hardfacing techniques have been widely used to enhance the resistance of surfaces against degradation mechanisms. Applying a surface coating improves wear and corrosion resistance and ensures reliability and long-term performance of coated parts. A protective layer or barrier on the components avoids the direct mechanical and chemical contacts of tool surfaces with process media and will reduce the material loss and ultimately its failure. Laser cladding as an advanced hardfacing technique has been widely used for industrial applications in order to develop a protective coating with desired material properties. During the laser cladding, coating material is fused into the base material by means of a laser beam in order to rebuild a damaged part's surface or to enhance its surface function. In the hardfacing techniques such as atmospheric plasma spraying (APS), high velocity oxygen-fuel (HVOF), and laser cladding, mixing of coating materials with underneath surface has to be minimized in order to utilize the properties of the coating material most effectively. In this regard, laser cladding offers advantages due to creating coating layers with superior properties in terms of purity, homogeneity, low dilution, hardness, bonding, and microstructure. In the development of modern materials for hardfacing applications, the functionality is often improved by combining materials with different properties into composites. Metal Matrix Composite (MMC) coating is a composite material with two constituent parts, i.e., matrix and the reinforcement. This class of composites are addressing improved mechanical properties such as stiffness, strength, toughness, and tribological and chemical resistance. Fabrication of MMCs is to achieve a combination of properties not achievable by any of the materials acting alone. MMCs have attracted significant attention for decades due to their combination of wear-resistivity, corrosion-resistivity, thermal, electrical and magnetic properties. Presently, there is a strong emphasis on the development of advanced functional coatings for corrosion, erosion, and wear protection for different industrial applications. In this research, a laser cladding system equipped with a high power direct diode laser associated with gas driven metal powder delivery system was used to develop advanced MMC coatings. The high power direct diode laser used in this study offers wider beam spot, shorter wavelength and uniform power distribution. These properties make the cladding set-up ideal for coating due to fewer cladding tracks, lower operation cost, higher laser absorption, and improved coating qualities. In order to prevent crack propagation, porosity, and uniform dispersion of carbides in MMC coating, cladding procedure was assisted by an induction heater as a second heat source. The developed defect free MMC coatings were combined with nano-size particles of WC, rare earth (RE) element (La2O3), and Mo as a refractory metal to enhance mechanical properties, chemical composition, and subsequently improve the tribological performance of the coatings. The resistance of developed MMC coatings were examined under highly accelerated slurry erosion, corrosion, and wear as the most frequently encountered failure modes of mechanical components. The microstructure, mechanical properties, and the level of induced residual stress on the coating after cladding procedure are closely related to cladding process variables. Study about the effect of processing parameters on clad quality and experienced thermal history and thermally-induced stress evolution requires both theoretical and experimental understanding of the associated physical phenomena. Numerical modeling offers a cost-efficient way to better understand the related complex physics in laser cladding process. It helps to reveal the effects and significance of each processing parameters on the desired characteristics of clad parts. Successful numerical simulation can provide unique insight into complex laser cladding process, efficiently calculate the complex procedure, and help to obtain coating parts with quality integrity. Therefore, current study develops a three-dimensional (3D) transient and uncoupled thermo-elastic-plastic model to study thermal history, molten pool evolution, thermally induced residual stress, and the effect of utilizing an induction heater as a second heat source on the mechanical properties and microstructural properties of final cladded coating.

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

Report on the BWR owners group radiation protection/ALARA Committee

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

Aldrich, L.R.

1995-03-01

Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements inmore » relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming {open_quotes}World Class{close_quotes} performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance.« less

Effect of laser power on clad metal in laser-TIG combined metal cladding

NASA Astrophysics Data System (ADS)

Utsumi, Akihiro; Hino, Takanori; Matsuda, Jun; Tasoda, Takashi; Yoneda, Masafumi; Katsumura, Munehide; Yano, Tetsuo; Araki, Takao

2003-03-01

TIG arc welding has been used to date as a method for clad welding of white metal as bearing material. We propose a new clad welding process that combines a CO2 laser and a TIG arc, as a method for cladding at high speed. We hypothesized that this method would permit appropriate control of the melted quantity of base metal by varying the laser power. We carried out cladding while varying the laser power, and investigated the structure near the boundary between the clad layer and the base metal. Using the laser-TIG combined cladding, we found we were able to control appropriately the degree of dilution with the base metal. By applying this result to subsequent cladding, we were able to obtain a clad layer of high quality, which was slightly diluted with the base metal.

Fabrication of a Porous Fiber Cladding Material Using Microsphere Templating for Improved Response Time with Fiber Optic Sensor Arrays

PubMed Central

Henning, Paul E.; Rigo, M. Veronica; Geissinger, Peter

2012-01-01

A highly porous optical-fiber cladding was developed for evanescent-wave fiber sensors, which contains sensor molecules, maintains guiding conditions in the optical fiber, and is suitable for sensing in aqueous environments. To make the cladding material (a poly(ethylene) glycol diacrylate (PEGDA) polymer) highly porous, a microsphere templating strategy was employed. The resulting pore network increases transport of the target analyte to the sensor molecules located in the cladding, which improves the sensor response time. This was demonstrated using fluorescein-based pH sensor molecules, which were covalently attached to the cladding material. Scanning electron microscopy was used to examine the structure of the templated polymer and the large network of interconnected pores. Fluorescence measurements showed a tenfold improvement in the response time for the templated polymer and a reliable pH response over a pH range of five to nine with an estimated accuracy of 0.08 pH units. PMID:22654644

Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

NASA Astrophysics Data System (ADS)

Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

2018-05-01

The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (< 1 dpa). Nonetheless, such zones undergo only nanoscopic swelling and a small hardness increase ( 10%), with no appreciable decrease in fracture strength. Thus, for this fluence and applied conditions, the integrity of the steel cladding is retained despite He2+ implantation.

Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

NASA Astrophysics Data System (ADS)

Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

2018-04-01

The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (< 1 dpa). Nonetheless, such zones undergo only nanoscopic swelling and a small hardness increase ( 10%), with no appreciable decrease in fracture strength. Thus, for this fluence and applied conditions, the integrity of the steel cladding is retained despite He2+ implantation.

Occurence and prediction of sigma phase in fuel cladding alloys for breeder reactors. [LMFBR

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

Anantatmula, R.P.

1982-01-01

In sodium-cooled fast reactor systems, fuel cladding materials will be exposed for several thousand hours to liquid sodium. Satisfactory performance of the materials depends in part on the sodium compatibility and phase stability of the materials. This paper mainly deals with the phase stability aspect, with particular emphasis on sigma phase formation of the cladding materials upon extended exposures to liquid sodium. A new method of predicting sigma phase formation is proposed for austenitic stainless steels and predictions are compared with the experimental results on fuel cladding materials. Excellent agreement is obtained between theory and experiment. The new method ismore » different from the empirical methods suggested for superalloys and does not suffer from the same drawbacks. The present method uses the Fe-Cr-Ni ternary phase diagram for predicting the sigma-forming tendencies and exhibits a wide range of applicability to austenitic stainless steels and heat-resistant Fe-Cr-Ni alloys.« less

Nuclear-powered pacemaker fuel cladding study. [Difficulty of dissolving cladding and /sup 238/PuO/sub 2/ for obtaining materials for acts of terrorism

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

Shoup, R.L.

1976-07-01

The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a /sup 238/PuO/sub 2/-powered pacemaker could be transformed into a terrorism weapon.

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

An analytical model to predict and minimize the residual stress of laser cladding process

NASA Astrophysics Data System (ADS)

Tamanna, N.; Crouch, R.; Kabir, I. R.; Naher, S.

2018-02-01

Laser cladding is one of the advanced thermal techniques used to repair or modify the surface properties of high-value components such as tools, military and aerospace parts. Unfortunately, tensile residual stresses generate in the thermally treated area of this process. This work focuses on to investigate the key factors for the formation of tensile residual stress and how to minimize it in the clad when using dissimilar substrate and clad materials. To predict the tensile residual stress, a one-dimensional analytical model has been adopted. Four cladding materials (Al2O3, TiC, TiO2, ZrO2) on the H13 tool steel substrate and a range of preheating temperatures of the substrate, from 300 to 1200 K, have been investigated. Thermal strain and Young's modulus are found to be the key factors of formation of tensile residual stresses. Additionally, it is found that using a preheating temperature of the substrate immediately before laser cladding showed the reduction of residual stress.

Metal clad aramid fibers for aerospace wire and cable

NASA Technical Reports Server (NTRS)

Tokarsky, Edward W.; Dunham, Michael G.; Hunt, James E.; Santoleri, E. David; Allen, David B.

1995-01-01

High strength light weight metal clad aramid fibers can provide significant weight savings when used to replace conventional metal wire in aerospace cable. An overview of metal clad aramid fiber materials and information on performance and use in braided electrical shielding and signal conductors is provided.

METHOD OF MANUFACTURE OF METAL ENCASED CORE MATERIAL

DOEpatents

Peters, E.J.

1963-06-01

A method of making reactor fuel elements in which the fissionable material is encased and sealed in steel or aluminum cladding having an enclosed channel from the fissionable material to the surface of the cladding at one end is described. Heat and pressure sufficient to bond the assembled fuel element are applied in a nonoxidizing atmosphere. (AEC)

Deoxyribonucleic acid (DNA) cladding layers for nonlinear-optic-polymer-based electro-optic devices

NASA Astrophysics Data System (ADS)

Grote, James G.; Ogata, Naoya; Diggs, Darnell E.; Hopkins, Frank K.

2003-07-01

Nonlinear optic (NLO) polymer based electro-optic devices have been achieving world record low half wave voltages and high frequencies over the last 2-3 years. Part of the advancement is through the use of relatively more conductive polymers for the cladding layers. Based on the current materials available for these cladding materials, however, the desired optical and electromagnetic properites are being balanced for materials processability. One does not want the solvent present in one layer to dissovle the one deposited underneath, or be dissolved by the one being deposited on top. Optimized polymer cladding materials, to further enhance device performance, are continuing to be investigated. Thin films of deoxyribonucleic acid (DNA), derived from salmon sperm, show promise in providing both the desired optical and magnetic properties, as well as the desired resistance to various solvents used for NLO polymer device fabrication. Thin films of DNA were deposited on glass and silicon substrates and the film quality, optical and electromagnetic properties and resistance to various solvents were characterized.

Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

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

Mickalonis, J. I.

2015-08-31

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material withmore » the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.« less

Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

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

Mickalonis, J. I.

2015-08-01

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material withmore » the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.« less

Development of Advanced Ods Ferritic Steels for Fast Reactor Fuel Cladding

NASA Astrophysics Data System (ADS)

Ukai, S.; Oono, N.; Ohtsuka, S.; Kaito, T.

Recent progress of the 9CrODS steel development is presented focusing on their microstructure control to improve sufficient high-temperature strength as well as cladding manufacturing capability. The martensitic 9CrODS steel is primarily candidate cladding materials for the Generation IV fast reactor fuel. They are the attractive composite-like materials consisting of the hard residual ferrite and soft tempered martensite, which are able to be easily controlled by α-γ phase transformation. The residual ferrite containing extremely nanosized oxide particles leads to significantly improved creep rupture strength in 9CrODS cladding. The creep strength stability at extended time of 60,000 h at 700 ºC is ascribed to the stable nanosized oxide particles. It was also reviewed that 9CrODS steel has well irradiation stability and fuel pin irradiation test was conducted up to 12 at% burnup and 51 dpa at the cladding temperature of 700ºC.

The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

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

Wang, Jy-An John; Jiang, Hao

2016-01-12

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at the Oak Ridge National Laboratory (ORNL) and is described fully in US Patent Application 20060070455, “Expanded plug method for developing circumferential mechanical properties of tubular materials.” This method is designed for testing fuel rod cladding ductility in a hot cell using an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are themore » simplicity of measuring the test component assembly in the hot cell and the direct measurement of the specimen’s strain. It was also found that cladding strength could be determined from the test results.« less

Fuel pin

DOEpatents

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

1987-11-24

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

Fuel pin

DOEpatents

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

1989-10-03

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

Fuel pin

DOEpatents

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

1989-01-01

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

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

Method of bonding

DOEpatents

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

1987-12-01

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

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

Brown, Nicholas; Burns, Joseph R.

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

Instant release fraction corrosion studies of commercial UO2 BWR spent nuclear fuel

NASA Astrophysics Data System (ADS)

Martínez-Torrents, Albert; Serrano-Purroy, Daniel; Sureda, Rosa; Casas, Ignasi; de Pablo, Joan

2017-05-01

The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel. The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

NASA Astrophysics Data System (ADS)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2017-12-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

NASA Astrophysics Data System (ADS)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2018-02-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Rectangular-cladding silicon slot waveguide with improved nonlinear performance

NASA Astrophysics Data System (ADS)

Huang, Zengzhi; Huang, Qingzhong; Wang, Yi; Xia, Jinsong

2018-04-01

Silicon slot waveguides have great potential in hybrid silicon integration to realize nonlinear optical applications. We propose a rectangular-cladding hybrid silicon slot waveguide. Simulation result shows that, with a rectangular-cladding, the slot waveguide can be formed by narrower silicon strips, so the two-photon absorption (TPA) loss in silicon is decreased. When the cladding material is a nonlinear polymer, the calculated TPA figure of merit (FOMTPA) is 4.4, close to the value of bulk nonlinear polymer of 5.0. This value confirms the good nonlinear performance of rectangular-cladding silicon slot waveguides.

Impact of nonabsorbing control rod tips on kinetics feedback for BWR turbine trip without bypass RETRAN-03 analyses

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

Feltus, M.A.; Knerr, R.; Shoop, U.

1993-01-01

RETRAN-03 studies were performed for the boiling water reactor (BWR) turbine trip without bypass (TTWOB) event to investigate how the non-neutron-absorbing material on control rod tips affect scram delay timing and reactivity feedback. Scram delay, Doppler temperature, and moderator void (density) feedback were varied to assess their relative impact on kinetics behavior. Although a generic point-kinetics RETRAN-03 TTWOB model 2 was employed, actual plant information was used to develop the basic and parametric cases.

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

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

Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been storedmore » on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.« less

Review and perspective: Sapphire optical fiber cladding development for harsh environment sensing

NASA Astrophysics Data System (ADS)

Chen, Hui; Buric, Michael; Ohodnicki, Paul R.; Nakano, Jinichiro; Liu, Bo; Chorpening, Benjamin T.

2018-03-01

The potential to use single-crystal sapphire optical fiber as an alternative to silica optical fibers for sensing in high-temperature, high-pressure, and chemically aggressive harsh environments has been recognized for several decades. A key technological barrier to the widespread deployment of harsh environment sensors constructed with sapphire optical fibers has been the lack of an optical cladding that is durable under these conditions. However, researchers have not yet succeeded in incorporating a high-temperature cladding process into the typical fabrication process for single-crystal sapphire fibers, which generally involves seed-initiated fiber growth from the molten oxide state. While a number of advances in fabrication of a cladding after fiber-growth have been made over the last four decades, none have successfully transitioned to a commercial manufacturing process. This paper reviews the various strategies and techniques for fabricating an optically clad sapphire fiber which have been proposed and explored in published research. The limitations of current approaches and future prospects for sapphire fiber cladding are discussed, including fabrication methods and materials. The aim is to provide an understanding of the past research into optical cladding of sapphire fibers and to assess possible material systems for future research on this challenging problem for harsh environment sensors.

77 FR 38825 - Clad Steel Plate From Japan; Scheduling of a Full Five-Year Review Concerning the Antidumping...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-29

... Japan; Scheduling of a Full Five-Year Review Concerning the Antidumping Duty Order on Clad Steel Plate From Japan AGENCY: United States International Trade Commission. ACTION: Notice. SUMMARY: The... order on clad steel plate from Japan would be likely to lead to continuation or recurrence of material...

Part Repairing Using A Hybrid Manufacturing System (Preprint)

DTIC Science & Technology

2007-03-01

laser . The laser processing parameters for cladding steel H13 powder were 600W with a stand-off distance from the nozzle to the top of the clad of 0.5...Journal of Materials Processing Technology, 2002:122, 63-68. [11]Richter, K., Orban, S., and Nowotny, S., Laser cladding of the titanium alloy TI6242...was used to repair the corroded steam generator tubes in nuclear plants [9], and turbine blades were repaired using the laser cladding process [10

Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

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

Morikawa, Yoshitake

1995-03-01

The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolantmore » system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.« less

Explosion Clad for Upstream Oil and Gas Equipment

NASA Astrophysics Data System (ADS)

Banker, John G.; Massarello, Jack; Pauly, Stephane

2011-01-01

Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO2 and/or H2S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

NASA Astrophysics Data System (ADS)

Carroll, Spencer

As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel performance code developed by PNNL and used by the Nuclear Regulatory Commission (NRC) as a licensing code for US reactors. FRAPCON will give insight into how these new fuel-cladding combinations will affect cladding hoop stress and help determine if the new materials are feasible for use in a reactor. To accurately simulate the interaction between the new materials, a soft pellet model that allows for stresses on the pellet to affect pellet deformation will have to be implemented. Currently, FRAPCON uses a rigid pellet model that does not allow for feedback of the cladding onto the pellet. Since SiC does not creep at the temperatures being considered and is not ductile, any PCMI create a much higher interfacial pressure than is possible with Zircaloy. Because of this, it is necessary to implement a model that allows for pellet creep to alleviate some of these cladding stresses. These results will then be compared to FEMAXI-6, a Japanese fuel performance code that already calculates pellet stress and allows for cladding feedback onto the pellet. This research is intended to be a continuation and verification of previous work done by USC on the analysis of accident tolerant fuels with alternative claddings and is intended to prove that a soft pellet model is necessary to accurately model any fuel with SiC cladding.

Methodology for Mechanical Property Testing on Fuel Cladding Using an Expanded Plug Wedge Test

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

Wang, Jy-An John; Jiang, Hao

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at ORNL and is described fully in US Patent Application 20060070455, Expanded plug method for developing circumferential mechanical properties of tubular materials. This method is designed for testing fuel rod cladding ductility in a hot cell utilizing an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the testmore » component assembly in the hot cell and the direct measurement of specimen strain. It was also found that cladding strength could be determined from the test results. The basic approach of this test method is to apply an axial compressive load to a cylindrical plug of polyurethane (or other materials) fitted inside a short ring of the test material to achieve radial expansion of the specimen. The diameter increase of the specimen is used to calculate the circumferential strain accrued during the test. The other two basic measurements are total applied load and amount of plug compression (extension). A simple procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves. However, several deficiencies exist in this expanded-plug loading ring test, which will impact accuracy of test results and introduce potential shear failure of the specimen due to inherited large axial compressive stress from the expansion plug test. First of all, the highly non-uniform stress and strain distribution resulted in the gage section of the clad. To ensure reliable testing and test repeatability, the potential for highly non-uniform stress distribution or displacement/strain deformation has to be eliminated at the gage section of the specimen. Second, significant compressive stresses were induced by clad bending deformation due to a clad bulging effect (or the barreling effect). The barreling effect caused very large localized shear stress in the clad and left testing material at a high risk of shear failure. The above combined effects will result in highly non-conservative predictions both in strength and ductility of the tested clad, and the associated mechanical properties as well. To overcome/mitigate the mentioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. Through detailed parameter investigation on specific geometry designs, careful filtering of material for the expansion plug, as well as adding newly designed parts to the testing system, a method to reconcile the potential non-conservatism embedded in the expansion plug test system has been discovered. A modified expansion plug testing protocol has been developed based on the method. In order to closely resemble thin-wall theory, a general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor called -factor is defined to correlate the ring load P into hoop stress . , = . The generated stress-strain curve agrees very well with tensile test data in both the elastic and plastic regions.« less

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

Wood, Elizabeth Sooby; Parker, Stephen Scott; Nelson, Andrew Thomas

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign is currently supporting a range of experimental efforts aimed at the development and qualification of ‘accident tolerant’ nuclear fuel forms. One route to enhance the accident tolerance of nuclear fuel is to replace the zirconium alloy cladding, which is prone to rapid oxidation in steam at elevated temperatures, with a more oxidation-resistant cladding. Several cladding replacement solutions have been envisaged. The cladding can be completely replaced with a more oxidation resistant alloy, a layered approach can be used to optimize the strength, creep resistance, and oxidation tolerance of various materials,more » or the existing zirconium alloy cladding can be coated with a more oxidation-resistant material. Molybdenum is one candidate cladding material favored due to its high temperature creep resistance. However, it performs poorly under autoclave testing and suffers degradation under high temperature steam oxidation exposure. Development of composite cladding architectures consisting of a molybdenum core shielded by a molybdenum disilicide (MoSi 2) coating is hypothesized to improve the performance of a Mo-based cladding system. MoSi 2 was identified based on its high temperature oxidation resistance in O 2 atmospheres (e.g. air and “wet air”). However, its behavior in H 2O is less known. This report presents thermogravimetric analysis (TGA), scanning electron microscopy (SEM), and x-ray diffraction (XRD) results for MoSi 2 exposed to 670-1498 K water vapor. Synthetic air (80-20%, Ar-O 2) exposures were also performed, and those results are presented here for a comparative analysis. It was determined that MoSi 2 displays drastically different oxidation behavior in water vapor than in dry air. In the 670-1498 K temperature range, four distinct behaviors are observed. Parabolic oxidation is exhibited in only 670-773 K water vapor, a temperature range in which the material pests in dry O 2 environments. From 877-1084 K in water vapor, MoSi 2 undergoes rapid mass gain resulting in oxidation throughout the bulk of the sample at 980 K and 1084 K. The resulting material displays swelling and warping after the 980-1084 K exposures. A pre-passivation heat treatment performed at 1395 K was found capable of producing a coarse SiO 2 layer that limited pesting at lower temperatures in water vapor over the time periods investigated.« less

Fiber optic moisture sensor

DOEpatents

Kirkham, R.R.

1984-08-03

A method and apparatus for sensing moisture changes by utilizing optical fiber technology. One embodiment uses a reflective target at the end of an optical fiber. The reflectance of the target varies with its moisture content and can be detected by a remote unit at the opposite end of the fiber. A second embodiment utilizes changes in light loss along the fiber length. This can be attributed to changes in reflectance of cladding material as a function of its moisture content. It can also be affected by holes or inserts interposed in the cladding material and/or fiber. Changing light levels can also be coupled from one fiber to another in an assembly of fibers as a function of varying moisture content in their overlapping lengths of cladding material.

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

DTIC Science & Technology

2013-06-01

Densities ............................................................................................................ 21 2.3 Fuel Mass (Core Total...70 7.1 Geometry, Material Density, and Mass Summary for All Cores...21 Table 3: Fuel Rod Masses for Different Clads

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.

Structural transformations in hull material clad by nitrogen stainless steel using various methods

NASA Astrophysics Data System (ADS)

Sagaradze, V. V.; Kataeva, N. V.; Mushnikova, S. Yu.; Khar'kov, O. A.; Kalinin, G. Yu.; Yampol'skii, V. D.

2014-02-01

Specimens of a 10N3KhDMBF shipbuilding hull steel were clad by a 04Kh20N6G11M2AFB nitrogen austenitic steel using various treatment conditions, which included hot rolling, austenitic facing, and explosive welding followed by hot rolling and heat treatment. Between the base and cladding materials, an intermediate layer with variable concentrations of chromium, manganese, and nickel was found, in which a martensitic structure was formed. In all the cases, the strength of bonding of the cladding layer to the hull steel (determined in tests for shear to fracture) was fairly high (σsh = 437-520 MPa). The only exception was the specimen produced by unidirectional facing without subsequent hot rolling (σsh = 308 MPa), in which nonfusions between the faced beads of stainless steel were detected.

Clad metals by roll bonding for SOFC interconnects

NASA Astrophysics Data System (ADS)

Chen, L.; Jha, B.; Yang, Zhenguo; Xia, Guang-Guang; Stevenson, Jeffry W.; Singh, Prabhakar

2006-08-01

High-temperature oxidation-resistant alloys are currently considered as a candidate material for construction of interconnects in intermediate-temperature solid oxide fuel cells. Among these alloys, however, different groups of alloys demonstrate different advantages and disadvantages, and few, if any, can completely satisfy the stringent requirements for the application. To integrate the advantages and avoid the disadvantages of different groups of alloys, cladding has been proposed as one approach in fabricating metallic layered interconnect structures. To examine the feasibility of this approach, the austenitic Ni-base alloy Haynes 230 and the ferritic stainless steel AL 453 were selected as examples and manufactured into a clad metal. Its suitability as an interconnect construction material was investigated. This paper provides a brief overview of the cladding approach and discusses the viability of this technology to fabricate the metallic layered-structure interconnects.

Laser cladding: repairing and manufacturing metal parts and tools

NASA Astrophysics Data System (ADS)

Sexton, Leo

2003-03-01

Laser cladding is presently used to repair high volume aerospace, automotive, marine, rail or general engineering components where excessive wear has occurred. It can also be used if a one-off high value component is either required or has been accidentally over-machined. The ultimate application of laser cladding is to build components up from nothing, using a laser cladding system and a 3D CAD drawing of the component. It is thus emerging that laser cladding can be classified as a special case of Rapid Prototyping (RP). Up to this point in time RP was seen, and is still seen, as in intermediately step between the design stage of a component and a finished working product. This can now be extended so that laser cladding makes RP a one-stop shop and the finished component is made from tool-steel or some alloy-base material. The marriage of laser cladding with RP is an interesting one and offers an alternative to traditional tool builders, re-manufacturers and injection mould design/repair industries. The aim of this paper is to discuss the emergence of this new technology, along with the transference of the process out of the laboratory and into the industrial workplace and show it is finding its rightful place in the manufacturing/repair sector. It will be shown that it can be used as a cost cutting, strategic material saver and consequently a green technology.

Effects of Thermal Aging on Material Properties, Stress Corrosion Cracking, and Fracture Toughness of AISI 316L Weld Metal

NASA Astrophysics Data System (ADS)

Lucas, Timothy; Forsström, Antti; Saukkonen, Tapio; Ballinger, Ronald; Hänninen, Hannu

2016-08-01

Thermal aging and consequent embrittlement of materials are ongoing issues in cast stainless steels, as well as duplex, and high-Cr ferritic stainless steels. Spinodal decomposition is largely responsible for the well-known "748 K (475 °C) embrittlement" that results in drastic reductions in ductility and toughness in these materials. This process is also operative in welds of either cast or wrought stainless steels where δ-ferrite is present. While the embrittlement can occur after several hundred hours of aging at 748 K (475 °C), the process is also operative at lower temperatures, at the 561 K (288 °C) operating temperature of a boiling water reactor (BWR), for example, where ductility reductions have been observed after several tens of thousands of hours of exposure. An experimental program was carried out in order to understand how spinodal decomposition may affect changes in material properties in Type 316L BWR piping weld metals. The study included material characterization, nanoindentation hardness, double-loop electrochemical potentiokinetic reactivation (DL-EPR), Charpy-V, tensile, SCC crack growth, and in situ fracture toughness testing as a function of δ-ferrite content, aging time, and temperature. SCC crack growth rates of Type 316L stainless steel weld metal under simulated BWR conditions showed an approximate 2 times increase in crack growth rate over that of the unaged as-welded material. In situ fracture toughness measurements indicate that environmental exposure can result in a reduction of toughness by up to 40 pct over the corresponding at-temperature air-tested values. Material characterization results suggest that spinodal decomposition is responsible for the degradation of material properties measured in air, and that degradation of the in situ properties may be a result of hydrogen absorbed during exposure to the high-temperature water environment.

Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

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

Asmolov, V.; Yegorova, L.; Kaplar, E.

1998-03-01

Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in thismore » paper.« less

Method and system for edge cladding of laser gain media

DOEpatents

Bayramian, Andrew James; Caird, John Allyn; Schaffers, Kathleen Irene

2014-03-25

A gain medium operable to amplify light at a gain wavelength and having reduced transverse ASE includes an input surface and an output surface opposing the input surface. The gain medium also includes a central region including gain material and extending between the input surface and the output surface along a longitudinal optical axis of the gain medium. The gain medium further includes an edge cladding region surrounding the central region and extending between the input surface and the output surface along the longitudinal optical axis of the gain medium. The edge cladding region includes the gain material and a dopant operable to absorb light at the gain wavelength.

In vitro testing of Nd:YAG laser processed calcium phosphate coatings.

PubMed

De Carlos, A; Lusquiños, F; Pou, J; León, B; Pérez-Amor, M; Driessens, F C M; Hing, K; Best, S; Bonfield, W

2006-11-01

Nd:YAG laser cladding is a new method for deposition of a calcium phosphate onto metallic surfaces of interest in implantology. The aim of this study was to compare the biologic response of MG-63 human osteoblast-like cells grown on Ti-6Al-4V substrates coated with a calcium phosphate layer applied using different methods: plasma spraying as reference material and Nd:YAG laser cladding as test material. Tissue culture polystyrene was used as negative control. The Nd:YAG laser clad material showed a behaviour similar to the reference material, plasma spray, respective to cell morphology (SEM observations), cell proliferation (AlamarBlue assay) and cytotoxicity of extracts (MTT assay). Proliferation, as measured by the AlamarBlue assay, showed little difference in the metabolic activity of the cells on the materials over an 18 day culture period. There were no significant differences in the cellular growth response on the test material when compared to the ones exhibited by the reference material. In the solvent extraction test all the extracts had some detrimental effect on cellular activity at 100% concentration, although cells incubated in the test material extract showed a proliferation rate similar to that of the reference material. To better understand the scope of these results it should be taken into account that the Nd:YAG clad coating has recently been developed. The fact that its in vitro performance is comparable to that produced by plasma spray, a material commercially available for more than ten years, indicates that this new laser based method could be of commercial interest in the near future.

Recycle of Zirconium from Used Nuclear Fuel Cladding: A Major Element of Waste Reduction

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

Collins, Emory D; DelCul, Guillermo D; Terekhov, Dmitri

2011-01-01

Feasibility tests were initiated to determine if the zirconium in commercial used nuclear fuel (UNF) cladding can be recovered in sufficient purity to permit re-use, and if the recovery process can be operated economically. Initial tests are being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Early results indicate that quantitative recovery can be accomplished and product contamination with alloy constituents can be controlled sufficiently to meet purification requirements. Future tests with actual radioactive UNF claddingmore » are planned. The objective of current research is to determine the feasibility of recovery and recycle of zirconium from used fuel cladding wastes. Zircaloy cladding, which contains 98+% of hafnium-free zirconium, is the second largest mass, on average {approx}25 wt %, of the components in used U.S. light-water-reactor fuel assemblies. Therefore, recovery and recycle of the zirconium would enable a large reduction in geologic waste disposal for advanced fuel cycles. Current practice is to compact or grout the cladding waste and store it for subsequent disposal in a geologic repository. This paper describes results of initial tests being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Future tests with actual radioactive UNF cladding are planned.« less

Fiber optic moisture sensor with moisture-absorbing reflective target

DOEpatents

Kirkham, Randy R.

1987-01-01

A method and apparatus for sensing moisture changes by utilizing optical fiber technology. One embodiment uses a reflective target at the end of an optical fiber. The reflectance of the target varies with its moisture content and can be detected by a remote unit at the opposite end of the fiber. A second embodiment utilizes changes in light loss along the fiber length. This can be attributed to changes in reflectance of cladding material as a function of its moisture content. It can also be affected by holes or inserts interposed in the cladding material and/or fiber. Changing light levels can also be coupled from one fiber to another in an assembly of fibers as a function of varying moisture content in their overlapping lengths of cladding material.

Towards AlN optical cladding layers for thermal management in hybrid lasers

NASA Astrophysics Data System (ADS)

Mathews, Ian; Lei, Shenghui; Nolan, Kevin; Levaufre, Guillaume; Shen, Alexandre; Duan, Guang-Hua; Corbett, Brian; Enright, Ryan

2015-06-01

Aluminium Nitride (AlN) is proposed as a dual function optical cladding and thermal spreading layer for hybrid ridge lasers, replacing current benzocyclobutene (BCB) encapsulation. A high thermal conductivity material placed in intimate contact with the Multi-Quantum Well active region of the laser allows rapid heat removal at source but places a number of constraints on material selection. AlN is considered the most suitable due to its high thermal conductivity when deposited at low deposition temperatures, similar co-efficient of thermal expansion to InP, its suitable refractive index and its dielectric nature. We have previously simulated the possible reduction in the thermal resistance of a hybrid ridge laser by replacing the BCB cladding material with a material of higher thermal conductivity of up to 319 W/mK. Towards this goal, we demonstrate AlN thin-films deposited by reactive DC magnetron sputtering on InP.

Efficient Geometry and Data Handling for Large-Scale Monte Carlo - Thermal-Hydraulics Coupling

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

Detailed coupling of thermal-hydraulics calculations to Monte Carlo reactor criticality calculations requires each axial layer of each fuel pin to be defined separately in the input to the Monte Carlo code in order to assign to each volume the temperature according to the result of the TH calculation, and if the volume contains coolant, also the density of the coolant. This leads to huge input files for even small systems. In this paper a methodology for dynamical assignment of temperatures with respect to cross section data is demonstrated to overcome this problem. The method is implemented in MCNP5. The method is verified for an infinite lattice with 3x3 BWR-type fuel pins with fuel, cladding and moderator/coolant explicitly modeled. For each pin 60 axial zones are considered with different temperatures and coolant densities. The results of the axial power distribution per fuel pin are compared to a standard MCNP5 run in which all 9x60 cells for fuel, cladding and coolant are explicitly defined and their respective temperatures determined from the TH calculation. Full agreement is obtained. For large-scale application the method is demonstrated for an infinite lattice with 17x17 PWR-type fuel assemblies with 25 rods replaced by guide tubes. Again all geometrical detailed is retained. The method was used in a procedure for coupled Monte Carlo and thermal-hydraulics iterations. Using an optimised iteration technique, convergence was obtained in 11 iteration steps.

Cladding burst behavior of Fe-based alloys under LOCA

DOE PAGES

Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; ...

2015-12-17

Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. Themore » most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.« less

ZIRCONIUM-CLADDING OF THORIUM

DOEpatents

Beaver, R.J.

1961-11-21

A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

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

Clad metals, roll bonding and their applications for SOFC interconnects

NASA Astrophysics Data System (ADS)

Chen, Lichun; Yang, Zhenguo; Jha, Bijendra; Xia, Guanguang; Stevenson, Jeffry W.

Metallic interconnects have been becoming an increasingly interesting topic in the development in intermediate temperature solid oxide fuel cells (SOFC). High temperature oxidation resistant alloys are currently considered as candidate materials. Among these alloys however, different groups of alloys demonstrate different advantages and disadvantages, and few if any can completely satisfy the stringent requirements for the application. To integrate the advantages and avoid the disadvantages of different groups of alloys, clad metal has been proposed for SOFC interconnect applications and interconnect structures. This paper gives a brief overview of the cladding approach and its applications, and discuss the viability of this technology to fabricate the metallic layered-structure interconnects. To examine the feasibility of this approach, the austenitic Ni-base alloy Haynes 230 and the ferritic stainless steel AL 453 were selected as examples and manufactured into a clad metal. Its suitability as an interconnect construction material was investigated.

In Situ Production of Hard Metal Matrix Composite Coating on Engineered Surfaces Using Laser Cladding Technique

NASA Astrophysics Data System (ADS)

Raza, Mohammad Shahid; Hussain, Manowar; Kumar, Vikash; Das, Alok Kumar

2017-01-01

The growing need for high wear-resistant surface with enhanced physical properties has led to extensive researches in the field of surface engineering. Laser cladding emerged to be a promising method to achieve these objectives in a cost-effective way. The present paper studies the viability of cladding of tungsten disulfide (WS2) powder by using 400 W continuous-wave fiber laser. WS2 was used as a coating material, which was decomposed at higher temperature and underwent several chemical reactions. By this process, in situ formation of metal matrix composites and hard face coating on the substrate surface were attained. The characterization of laser cladded surface was done to study its morphological, microstructural, mechanical and tribological properties. It was observed that cladding of WS2 powder on 304 SS resulted in the formation of Cr-W-C-Fe metal matrix composite having improved mechanical and tribological properties. The value of microhardness of the coated surface was found to increase three to four times in comparison with the parent material surface. Wear test results indicated a decrease in wear by 1/9th (maximum) as compared to the parent 304 SS surface. The volume fractions of tungsten particles on the cladded surface were also investigated through EDS analysis.

Radiation field control at the latest BWR plants -- design principle, operational experience and future subjects

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

Uchida, Shunsuke; Ohsumi, Katsumi; Takashima, Yoshie

1995-03-01

Improvements of operational procedures to control water chemistry, e.g., nickel/iron control, as well as application of hardware improvements for reducing radioactive corrosion products resulted in an extremely low occupational exposure of less than 0.5 man.Sv/yr without any serious impact on the radwaste system, for BWR plants involved in the Japanese Improvement and Standardization Program. Recently, {sup 60}C radioactively in the reactor water has been increasing due to less crud fixation on the two smooth surfaces of new type high performance fuels and to the pH drop caused by chromium oxide anions released from stainless steel structures and pipings. This increasemore » must be limited by changes in water chemistry, e.g., applications of modified nickel/iron ratio control and weak alkali control. Controlled water chemistry to optimize three points, the plant radiation level and integrities of fuel and structural materials, is the primary future subject for BWR water chemistry.« less

Hydrogen permeation in FeCrAl alloys for LWR cladding application

NASA Astrophysics Data System (ADS)

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

2015-06-01

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

Migration of Point Defects in the Field of a Temperature Gradient

NASA Astrophysics Data System (ADS)

Kozlov, A. V.; Portnykh, I. A.; Pastukhov, V. I.

2018-04-01

The influence of the temperature gradient over the thickness of the cladding of a fuel element of a fast-neutron reactor on the migration of point defects formed in the cladding material due to neutron irradiation has been studied. It has been shown that, under the action of the temperature gradient, the flux of vacancies onto the inner surface of the cladding is higher than the flux of interstitial atoms, which leads to the formation of a specific concentration profile in the cladding with a vacancy-depleted zone near the inner surface. The experimental results on the spatial distribution of pores over the cladding thickness have been presented with which the data on the concentration profiles and vacancy fluxes have been compared.

Modeling of Heat Transfer and Fluid Flow in the Laser Multilayered Cladding Process

NASA Astrophysics Data System (ADS)

Kong, Fanrong; Kovacevic, Radovan

2010-12-01

The current work examines the heat-and-mass transfer process in the laser multilayered cladding of H13 tool steel powder by numerical modeling and experimental validation. A multiphase transient model is developed to investigate the evolution of the temperature field and flow velocity of the liquid phase in the molten pool. The solid region of the substrate and solidified clad, the liquid region of the melted clad material, and the gas region of the surrounding air are included. In this model, a level-set method is used to track the free surface motion of the molten pool with the powder material feeding and scanning of the laser beam. An enthalpy-porosity approach is applied to deal with the solidification and melting that occurs in the cladding process. Moreover, the laser heat input and heat losses from the forced convection and heat radiation that occurs on the top surface of the deposited layer are incorporated into the source term of the governing equations. The effects of the laser power, scanning speed, and powder-feed rate on the dilution and height of the multilayered clad are investigated based on the numerical model and experimental measurements. The results show that an increase of the laser power and powder feed rate, or a reduction of the scanning speed, can increase the clad height and directly influence the remelted depth of each layer of deposition. The numerical results have a qualitative agreement with the experimental measurements.

History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

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

Larry Zirker; Nathan Jerred; Dr. Indrajit Charit

2012-03-01

Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.

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)

Removal of Sb-125 and Tc-99 from Liquid Radwaste by Novel Adsorbents

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

Harjula, R.O.; Koivula, R.; Paajanen, A.

2006-07-01

Novel proprietary metal oxide materials (MOM) have been tested for the removal of Sb-125 from simulated Floor Drain Waters of BWR. Antimony was present in the solutions as oxidized anionic form. Long term column experiment with simulated liquid that showed high Sb-125 removal at least up to 8000 bed volumes. One column experiments was carried out using nonradioactive Sb to exhaust the column. Leaching tests with 1000 ppm boric acid showed that 100 % of absorbed Sb remains in the sorbent material. Column experiments with real Fuel Pond Water from Olkiluoto NPP (BWR) showed reduction of Sb-125 (feed level 400more » Bq/L, 1.10{sup -5} {mu}Ci/mL) below detection limit (MDA = 1.7 Bq/L, 5.10{sup -8},{mu}Ci/mL). Additional experiments have also been carried out with pertechnetate (Tc-99) ions. Results indicate that MOM materials are efficient also for the removal of Tc-99 from concentrated NaNO{sub 3} solution. (authors)« less

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.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

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

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. CASL has endeavored to improve upon this approach by incorporating a microstructurally-based, atomistically-informed, zirconium alloy mechanical deformation analysis capability into the BISON-CASL engineering scale fuel performance code. Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed bymore » Lebensohn and Tome´ [2], has been coupled with BISON-CASL to represent the mechanistic material processes controlling the deformation behavior of the cladding. A critical component of VPSC is the representation of the crystallographic orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON-CASL and provides initial results utilizing the coupled functionality.« less

Studies of Lanthanide Transport in Metallic Fuel

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

Zhang, Jinsuo; Taylor, Christopher

Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity.

Fuel pin cladding

DOEpatents

Vaidyanathan, Swaminathan; Adamson, Martyn G.

1986-01-01

An improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

Numerical and Analytical Modeling of Laser Deposition with Preheating (Preprint)

DTIC Science & Technology

2007-03-01

temperature materials, Numerical Heat Transfer 11 (1987) 477-491. [9] L. Han, F.W. Liou, K.M. Phatk, Modeling of laser cladding with powder injection... cladding process. This laser additive manufacturing technique allows quick fabrication of fully-dense metallic components directly from Computer...1, laser deposition uses a focused laser beam as a heat source to create a melt pool on an underlying substrate. Powder material is then injected

(Project 14-6770) An Investigation to Establish Multiphysical Property Dataset of Nuclear Materials Based on in-situ Observations and Measurements

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

Tomar, Vikas; Haque, Aman; Hattar, Khalid

In-core nuclear materials including fuel pins and cladding materials fail due to issues including corrosion, mechanical wear, and pellet cladding interaction. In most such scenario microstructure dependent and corrosioninduced chemistry dependent property changes significantly affect performance of cladding, pellet, and housing. Emphasis of this work was on replace conventional pellet-cladding material models with a new straingradient viscoplasticity model that is informed by transmission electron microscopy (TEM) based measurements and by nanomechanical Raman spectroscopy (NMRS) based measurements. The TEM measurements are quantitative in nature and therefore reveal stress-strain relations with simultaneous insights into mechanisms of deformation at nanoscale. The NMRS measurementsmore » reveal the similar information at mesoscale along with additional information on relating local microstructural stresses with applied stresses. The resulting information is used to fit constants in the strain gradient viscoplasticity model as well as to validate one. During TEM measurements, a micro-electro-mechanical system based setup was developed with mechanical actuation, sensing, heating, and electrical loading. Contrary to post-mortem analysis or qualitative visualization, this setup combines direct visualization of the mechanisms behind deformation with measurement of stress, strain, thermal and electrical properties. The unique research philosophy of visualizing the microstructure at high resolution while measuring the properties led to fundamental understanding in grain size and temperature effects on measured mechanical properties such as fracture toughness. A key contribution is the role of mechanical loading boundary conditions to deconvolute the insitu TEM based nanoscale and NMRS based mesoscale data to bulk behavior. First the literature based pellet cladding mechanical interaction model based on the work of Retel’s and Williamson’s in literature work to predict tempurature and stress distribution in cladding and pellet at normal operating condition was analyzed. Later the data was fitted to find constants for a viscoplastic strain gradient model. The developed model still needs to be refined and calibrated using various experimental results. That remains the focus of future work. Overall, a major thrust of the work was therefore on active control of the microstructure (grain size, defect density and types) exploiting the multi-physics coupling in materials. In particular, using experiments the synergy of current density, mechanical stress and temperature were studied to annihilate defects and recrystallize grains. The developed model is being examined for implementation in BISON. Multiple invited talks, international journal publications, and conference publications were performed by students supported on this work. Another output is support multiple PhD and masters thesis students who will be an important asset for future basic nuclear research. Future Work Recommendations: A nuclear reactor operates under significant variations of thermal loads due to energy cycling and mechanical loads due to constraint effects. Significant thermal and chemical diffusion takes place at the pallet-cladding level. While the proposed work established new experimental approach and new dataset for Zircaloy-4, the irradiation level was in the range of 1-2 dpa. Samples with higher dpa need to be examined. Therefore, a continual of support of the performed work is essential. Currently, these are the only experiments that can measure the produced data. The work also needs to be extended to different fuel types and cladding types such as SiC and FeCrAl based claddings. A combination of datasets for these materials can then be used to analyze accurately predict behavior of critical pellet cladding systems in accident scenario with high heat flux and high thermal loads. This is a BIG unknown as if now.« less

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1980-04-29

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

Optimization of cladding parameters for resisting corrosion on low carbon steels using simulated annealing algorithm

NASA Astrophysics Data System (ADS)

Balan, A. V.; Shivasankaran, N.; Magibalan, S.

2018-04-01

Low carbon steels used in chemical industries are frequently affected by corrosion. Cladding is a surfacing process used for depositing a thick layer of filler metal in a highly corrosive materials to achieve corrosion resistance. Flux cored arc welding (FCAW) is preferred in cladding process due to its augmented efficiency and higher deposition rate. In this cladding process, the effect of corrosion can be minimized by controlling the output responses such as minimizing dilution, penetration and maximizing bead width, reinforcement and ferrite number. This paper deals with the multi-objective optimization of flux cored arc welding responses by controlling the process parameters such as wire feed rate, welding speed, Nozzle to plate distance, welding gun angle for super duplex stainless steel material using simulated annealing technique. Regression equation has been developed and validated using ANOVA technique. The multi-objective optimization of weld bead parameters was carried out using simulated annealing to obtain optimum bead geometry for reducing corrosion. The potentiodynamic polarization test reveals the balanced formation of fine particles of ferrite and autenite content with desensitized nature of the microstructure in the optimized clad bead.

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

DOE PAGES

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; ...

2016-07-15

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

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

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Finite Element Analysis of Laser Engineered Net Shape (LENS™) Tungsten Clad Squeeze Pins

NASA Astrophysics Data System (ADS)

Sakhuja, Amit; Brevick, Jerald R.

2004-06-01

In the aluminum high-pressure die-casting and indirect squeeze casting processes, local "squeeze" pins are often used to minimize internal solidification shrinkage in heavy casting sections. Squeeze pins frequently fail in service due to molten aluminum adhering to the H13 tool steel pins ("soldering"). A wide variety of coating materials and methods have been developed to minimize soldering on H13. However, these coatings are typically very thin, and experience has shown their performance on squeeze pins is highly variable. The LENS™ process was employed in this research to deposit a relatively thick tungsten cladding on squeeze pins. An advantage of this process was that the process parameters could be precisely controlled in order to produce a satisfactory cladding. Two fixtures were designed and constructed to enable the end and outer diameter (OD) of the squeeze pins to be clad. Analyses were performed on the clad pins to evaluate the microstructure and chemical composition of the tungsten cladding and the cladding-H13 substrate interface. A thermo-mechanical finite element analysis (FEA) was performed to assess the stress distribution as a function of cladding thickness on the pins during a typical casting thermal cycle. FEA results were validated via a physical test, where the clad squeeze pins were immersed into molten aluminum. Pins subjected to the test were evaluated for thermally induced cracking and resistance to soldering of the tungsten cladding.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

DOE PAGES

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; ...

2015-03-19

FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory insidemore » the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.« less

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

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

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less

(Boiling water reactor (BWR) CORA experiments)

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

Ott, L.J.

To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of themore » BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.« less

Cladding and duct materials for advanced nuclear recycle reactors

NASA Astrophysics Data System (ADS)

Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

2008-01-01

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

DOE PAGES

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.; ...

2016-12-01

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

Direct Metal Deposition of H13 Tool Steel on Copper Alloy Substrate: Parametric Investigation

NASA Astrophysics Data System (ADS)

Imran, M. Khalid; Masood, S. H.; Brandt, Milan

2015-12-01

Over the past decade, researchers have demonstrated interest in tribology and prototyping by the laser aided material deposition process. Laser aided direct metal deposition (DMD) enables the formation of a uniform clad by melting the powder to form desired component from metal powder materials. In this research H13 tool steel has been used to clad on a copper alloy substrate using DMD. The effects of laser parameters on the quality of DMD deposited clad have been investigated and acceptable processing parameters have been determined largely through trial-and-error approaches. The relationships between DMD process parameters and the product characteristics such as porosity, micro-cracks and microhardness have been analysed using scanning electron microscope (SEM), image analysis software (ImageJ) and microhardness tester. It has been found that DMD parameters such as laser power, powder mass flow rate, feed rate and focus size have an important role in clad quality and crack formation.

The effect of particle size on the heat affected zone during laser cladding of Ni-Cr-Si-B alloy on C45 carbon steel

NASA Astrophysics Data System (ADS)

Tanigawa, Daichi; Abe, Nobuyuki; Tsukamoto, Masahiro; Hayashi, Yoshihiko; Yamazaki, Hiroyuki; Tatsumi, Yoshihiro; Yoneyama, Mikio

2018-02-01

Laser cladding is one of the most useful surface coating methods for improving the wear and corrosion resistance of material surfaces. Although the heat input associated with laser cladding is small, a heat affected zone (HAZ) is still generated within the substrate because this is a thermal process. In order to reduce the area of the HAZ, the heat input must therefore be reduced. In the present study, we examined the effects of the powdered raw material particle size on the heat input and the extent of the HAZ during powder bed laser cladding. Ni-Cr-Si-B alloy layers were produced on C45 carbon steel substrates in conjunction with alloy powders having average particle sizes of 30, 40 and 55 μm, while measuring the HAZ area by optical microscopy. The heat input required for layer formation was found to decrease as smaller particles were used, such that the HAZ area was also reduced.

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

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

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

Development of new ferritic steels as cladding material for metallic fuel fast breeder reactor

NASA Astrophysics Data System (ADS)

Tokiwai, Moriyasu; Horie, Masaaki; Kako, Kenji; Fujiwara, Masayuki

1993-09-01

The excellent thermal, chemical and neutronic properties of metallic fuel (U-Pu-Zr alloy) will lead to drastic improvements in fast reactor safety and the related fuel cycle economy. Some new high molybdenum 12Cr ferritic stainless steel candidate cladding alloys have been designed to achieve the mechanical properties required for high performance metallic fuel elements. These candidate claddings were irradiated by ion bombardment and tested to determine their strength and creep rupture properties. A 12Cr-8Mo and a 12Cr-8Mo-0.1Y 2O 3 steel were fabricated into cladding via a powder metallurgy process and by a mechanical alloying process, respectively. These claddings had two and three times the creep rupture strength (pressurized at 650°C for 10000 h) of a conventional 12Cr ferritic steel (HT-9). These two steels also showed no void formation up to 350 dpa by Ni 3+ irradiation. A zircaloy-2 lined steel cladding tube has also been fabricated for the purpose of reducing fuel-cladding interdiffusion and chemical interaction.

Microfabricated bragg waveguide

DOEpatents

Fleming, James G.; Lin, Shawn-Yu; Hadley, G. Ronald

2004-10-19

A microfabricated Bragg waveguide of semiconductor-compatible material having a hollow core and a multilayer dielectric cladding can be fabricated by integrated circuit technologies. The microfabricated Bragg waveguide can comprise a hollow channel waveguide or a hollow fiber. The Bragg fiber can be fabricated by coating a sacrificial mandrel or mold with alternating layers of high- and low-refractive-index dielectric materials and then removing the mandrel or mold to leave a hollow tube with a multilayer dielectric cladding. The Bragg channel waveguide can be fabricated by forming a trench embedded in a substrate and coating the inner wall of the trench with a multilayer dielectric cladding. The thicknesses of the alternating layers can be selected to satisfy the condition for minimum radiation loss of the guided wave.

Fuel pin cladding

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1986-01-28

Disclosed is an improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients. 2 figs.

Fuel pin cladding

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1983-12-16

An improved fuel pin cladding, particularly adapted for use in breeder reactors, is described which consist of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel an/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

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

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

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

Wirth, Brian; Terrani, Kurt A.; Sweet, Ryan T.

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling.more » As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations of the IFA-796 rod 4 experiment, but show that varying the creep model (within the range investigated here) only provide a minimal increase in the fuel radius and maximum cladding hoop stress. Continued investigation of fuel behavioral models will include benchmarking the modified fuel creep model against available experimental data, as well as an investigation of the role that fuel cracking will play in the compliance of the fuel. Correctly calculating stress evolution in the fuel is key to assessing fuel behavior up to gap closure and the subsequent deformation of the cladding due to PCMI. The inclusion of frictional contact should also be investigated to determine the axial elongation of the fuel rods for comparison with data from this experiment.« less

Surface modification techniques for increased corrosion tolerance of zirconium fuel cladding

NASA Astrophysics Data System (ADS)

Carr, James Patrick, IV

Corrosion is a major issue in applications involving materials in normal and severe environments, especially when it involves corrosive fluids, high temperatures, and radiation. Left unaddressed, corrosion can lead to catastrophic failures, resulting in economic and environmental liabilities. In nuclear applications, where metals and alloys, such as steel and zirconium, are extensively employed inside and outside of the nuclear reactor, corrosion accelerated by high temperatures, neutron radiation, and corrosive atmospheres, corrosion becomes even more concerning. The objectives of this research are to study and develop surface modification techniques to protect zirconium cladding by the incorporation of a specific barrier coating, and to understand the issues related to the compatibility of the coatings examined in this work. The final goal of this study is to recommend a coating and process that can be scaled-up for the consideration of manufacturing and economic limits. This dissertation study builds on previous accident tolerant fuel cladding research, but is unique in that advanced corrosion methods are tested and considerations for implementation by industry are practiced and discussed. This work will introduce unique studies involving the materials and methods for accident tolerant fuel cladding research by developing, demonstrating, and considering materials and processes for modifying the surface of zircaloy fuel cladding. This innovative research suggests that improvements in the technique to modify the surface of zirconium fuel cladding are likely. Three elements selected for the investigation of their compatibility on zircaloy fuel cladding are aluminum, silicon, and chromium. These materials are also currently being investigated at other labs as alternate alloys and coatings for accident tolerant fuel cladding. This dissertation also investigates the compatibility of these three elements as surface modifiers, by comparing their microstructural and mechanical properties. To test their application for use in corrosive atmospheres, the corrosion behaviors are also compared in steam, water, and boric-acid environments. Various methods of surface modification were attempted in this investigation, including dip coating, diffusion bonding, casting, sputtering, and evaporation. The benefits and drawbacks of each method are discussed with respect to manufacturing and economic limits. Characterization techniques utilized in this work include optical microscopy, scanning electron microscopy, energy-dispersive spectroscopy, X-ray diffraction, nanoindentation, adhesion testing, and atomic force microscopy. The composition, microstructure, hardness, modulus, and coating adhesion were studied to provide encompassing properties to determine suitable comparisons and to choose an ideal method to scale to industrial applications. The experiments, results, and detailed discussions are presented in the following chapters of this dissertation research.

Laser Cladding of TiAl Intermetallic Alloy on Ti6Al4V -Process Optimization and Properties

NASA Astrophysics Data System (ADS)

Cárcel, B.; Serrano, A.; Zambrano, J.; Amigó, V.; Cárcel, A. C.

In order to improve Ti6Al4V high-temperature resistance and its tribological properties, the deposition of TiAl intermetallic (Ti-48Al-2Cr-2Nb) coating on a Ti6Al4V substrate by coaxial laser cladding has been investigated. Laser cladding by powder injection is an emerging laser material processing technique that allows the deposition of thick protective coatings on substrates,using a high power laser beam as heat source. Laser cladding is a multiple-parameter-dependent process. The main process parameters involved (laser power, powder feeding rate, scanning speed and preheating temperature) has been optimized. The microstructure and geometrical quantities (clad area and dilution) of the coating was characterized by optical microscopy and scanning electron microscopy (SEM). In addition the cooling rate of the clad during the process was measured by a dual-color pyrometer. This result has been related to defectology and mechanical coating properties.

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.

Mechanical behavior of aluminum-bearing ferritic alloys for accident-tolerant fuel cladding applications

NASA Astrophysics Data System (ADS)

Guria, Ankan

Nuclear power currently provides about 13% of electrical power worldwide. Nuclear reactors generating this power traditionally use Zirconium (Zr) based alloys as the fuel cladding material. Exothermic reaction of Zr with steam under accident conditions may lead to production of hydrogen with the possibility of catastrophic consequences. Following the Fukushima-Daiichi incident, the exploration of accident-tolerant fuel cladding materials accelerated. Aluminum-rich (around 5 wt. %) ferritic steels such as Fecralloy, APMT(TM) and APM(TM) are considered as potential materials for accident-tolerant fuel cladding applications. These materials create an aluminum-based oxide scale protecting the alloy at elevated temperatures. Tensile deformation behavior of the above alloys was studied at different temperatures (25-500 °C) at a strain rate of 10-3 s-1 and correlated with microstructural characteristics. Higher strength and decent ductility of APMT(TM) led to further investigation of the alloy at various combination of strain rates and temperatures followed by fractography and detailed microscopic analyses. Serrations appeared in the stress-strain curves of APMT(TM) and Fecralloy steel tested in a limited temperature range (250-400 °C). The appearance of serrations is explained on the basis of dynamic strain aging (DSA) effect due to solute-dislocation interactions. The research in this study is being performed using the funds received from the US DOE Office of Nuclear Energy's Nuclear Energy University Programs (NEUP).

Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

NASA Astrophysics Data System (ADS)

Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

2018-02-01

SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

Propagation of Electromagnetic Waves in Slab Waveguide Structure Consisting of Chiral Nihility Claddings and Negative-Index Material Core Layer

NASA Astrophysics Data System (ADS)

Helal, Alaa N. Abu; Taya, Sofyan A.; Elwasife, Khitam Y.

2018-06-01

The dispersion equation of an asymmetric three-layer slab waveguide, in which all layers are chiral materials is presented. Then, the dispersion equation of a symmetric slab waveguide, in which the claddings are chiral materials and the core layer is negative index material, is derived. Normalized cut-off frequencies, field profile, and energies flow of right-handed and left-handed circularly polarized modes are derived and plotted. We consider both odd and even guided modes. Numerical results of guided low-order modes are provided. Some novel features, such as abnormal dispersion curves, are found.

Porous Materials with Ultralow Optical Constants for Integrated Optical Device Applications

NASA Astrophysics Data System (ADS)

Chen, Hsuen-Li; Hsieh, Chung-I; Cheng, Chao-Chia; Chang, Chia-Pin; Hsu, Wen-Hau; Wang, Way-Seen; Liu, Po-Tsun

2005-07-01

Ultralow dielectric constant (<2.0) porous materials have received much attention as next-generation dielectric materials. In this study, optical properties of porous-methyl-silsesquioxane(MSQ)-like films (porous polysilazane, PPSZ) were characterized for optical waveguide devices applications. Measured results indicate that the refractive index is decreased to approximately 1.320 as the hydration time exceeds 24 h. The measured refractive index is about 1.163 at a wavelength of 1550 nm. PPSZ films have low absorption in the 500 to 2000 nm wavelength regime. Because of their relatively low refractive index and low absorption over a large spectral regime, PPSZ films can be good cladding materials for use in optically integrated devices with many high-refractive-index materials such as silicon oxide, silicon nitride, silicon, and polymers. We demonstrate two structures, ridge waveguides and large-angle Y-branch power splitters, composed of PPSZ and SU8 films to illustrate the use of low dielectric constant (K) cladding materials. The simulation results indicate that the PPSZ films provide better confinement of light. Experimentally, a large-angle Y-branch power splitter with PPSZ cladding can be used to guide waves with the large branching angle of 33.58°.

Apparatus and method for critical current measurements

DOEpatents

Martin, Joe A.; Dye, Robert C.

1992-01-01

An apparatus for the measurement of the critical current of a superconductive sample, e.g., a clad superconductive sample, the apparatus including a conductive coil, a means for maintaining the coil in proximity to a superconductive sample, an electrical connection means for passing a low amplitude alternating current through the coil, a cooling means for maintaining the superconductive sample at a preselected temperature, a means for passing a current through the superconductive sample, and, a means for monitoring reactance of the coil, is disclosed, together with a process of measuring the critical current of a superconductive material, e.g., a clad superconductive material, by placing a superconductive material into the vicinity of the conductive coil of such an apparatus, cooling the superconductive material to a preselected temperature, passing a low amplitude alternating current through the coil, the alternating current capable of generating a magnetic field sufficient to penetrate, e.g., any cladding, and to induce eddy currents in the superconductive material, passing a steadily increasing current through the superconductive material, the current characterized as having a different frequency than the alternating current, and, monitoring the reactance of the coil with a phase sensitive detector as the current passed through the superconductive material is steadily increased whereby critical current of the superconductive material can be observed as the point whereat a component of impedance deviates.

78 FR 7451 - Clad Steel Plate From Japan; Determination

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-01

... Japan; Determination On the basis of the record \\1\\ developed in the subject five-year review, the... from Japan would be likely to lead to continuation or recurrence of material injury to an industry in... USITC Publication 4370 (January 2013), entitled Clad Steel Plate from Japan: Investigation No. 731-TA...

PROCESS OF FORMING POWDERED MATERIAL

DOEpatents

Glatter, J.; Schaner, B.E.

1961-07-14

A process of forming high-density compacts of a powdered ceramic material is described by agglomerating the powdered ceramic material with a heat- decompossble binder, adding a heat-decompossble lubricant to the agglomerated material, placing a quantity of the material into a die cavity, pressing the material to form a compact, pretreating the compacts in a nonoxidizing atmosphere to remove the binder and lubricant, and sintering the compacts. When this process is used for making nuclear reactor fuel elements, the ceramic material is an oxide powder of a fissionsble material and after forming, the compacts are placed in a cladding tube which is closed at its ends by vapor tight end caps, so that the sintered compacts are held in close contact with each other and with the interior wall of the cladding tube.

M3FT-16OR0203052-Test Design for FeCrAl Alloy Tube Irradiation in HFIR

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

Terrani, Kurt A.; Petrie, Christian M.

2016-05-01

This calculation summarizes thermal analyses of a flexible rabbit design for irradiating a variety of pressurized water reactor (PWR) cladding materials (stainless steel, iron-chromium aluminum [FeCrAl], Zircaloy, and Inconel) with variable dimensions at a temperature of 350 °C in the flux trap of the High Flux Isotope Reactor (HFIR). The design can accommodate standard cladding for outer diameters (ODs) of approximately 9.50 mm with thickness ranging from 0.30 mm to 0.70 mm. The length is generally between 10 and 50 mm. The specimens contain moly inserts with a variable OD that provides the heat flux necessary to achieve the designmore » temperature with such a small fixed gas gap. The primary outer containment is an Al-6061 housing with a slightly enlarged inner diameter (ID) of 9.60 mm. The specimen temperature is controlled by determining a helium/argon gas mixture specific to the as-built specimen and housing. Variables that affect the required gas mixture are the cladding material (thermal expansion, density, heat generation rate), cladding OD, housing ID, and cladding ID. This calculation documents the analyses performed to determine required gas mixtures for a variety of scenarios.« less

Review of CTF s Fuel Rod Modeling Using FRAPCON-4.0 s Centerline Temperature Predictions

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

Toptan, Aysenur; Salko, Robert K; Avramova, Maria

Coolant Boiling in Rod Arrays Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF s fuel rod modeling to address shortcomings in CTF s temperature predictions. CTF is compared to FRAPCON which is U.S. NRC s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes inmore » fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF s dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF s dynamic gap conductance model. Improving the CTF s dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.« less

High-Temperature Tolerance in Multi-Scale Cermet Solar-Selective Absorbing Coatings Prepared by Laser Cladding.

PubMed

Pang, Xuming; Wei, Qian; Zhou, Jianxin; Ma, Huiyang

2018-06-19

In order to achieve cermet-based solar absorber coatings with long-term thermal stability at high temperatures, a novel single-layer, multi-scale TiC-Ni/Mo cermet coating was first prepared using laser cladding technology in atmosphere. The results show that the optical properties of the cermet coatings using laser cladding were much better than the preplaced coating. In addition, the thermal stability of the optical properties for the laser cladding coating were excellent after annealing at 650 °C for 200 h. The solar absorptance and thermal emittance of multi-scale cermet coating were 85% and 4.7% at 650 °C. The results show that multi-scale cermet materials are more suitable for solar-selective absorbing coating. In addition, laser cladding is a new technology that can be used for the preparation of spectrally-selective coatings.

Holmium Doped Solid State Laser Resonantly Pumped and Q-Switched by Novel GaSb-Based Photonic Devices

DTIC Science & Technology

2011-08-31

increased overlap with p-cladding, presumably due to dominant role of inter valence band absorption [7]. Details of the conduction band structure of the...absorption to total loss. In the specific structures used here the n-cladding composition resulted into material with three valleys in conduction band to...materials. The beam properties of the high power 2 μm emitting GaSb -based diode lasers was improved by utilization of the waveguide structure with

Planar polymer waveguides with a graded-index profile resulting from intermixing of methacrylates in closed microchannels

NASA Astrophysics Data System (ADS)

Missinne, Jeroen; Misseeuw, Lara; Liu, Xiang; Salter, Patrick S.; Van Steenberge, Geert; Adesanya, Kehinde; Van Vlierberghe, Sandra; Booth, Martin J.; Dubruel, Peter

2018-02-01

Graded-index waveguides are known to exhibit lower losses and considerably larger bandwidths compared to step-index waveguides. The present work reports on a new concept for realizing such waveguides on a planar substrate by capillary filling microchannels (cladding) with monomer solution (core). A graded-index profile is obtained by intermixing between the core and cladding material at the microchannel interface. To this end, various ratios of methyl methacrylate (MMA) and octafluoropentyl methacrylate (OFPMA) were evaluated as starting monomers and the results showed that the polymers P50:50 (50:50 MMA:OFPMA) and P0:100 (100% OFPMA) were suitable to be applied as waveguide core and cladding material respectively. Light guiding in the resulting P50:50/P0:100 waveguides was demonstrated and the refractive-index profile was quantified and compared with that of conventional step-index waveguides. The results for both cases were clearly different and a gradual refractive index transition between the core and cladding was found for the newly developed waveguides. Although the concept has been demonstrated in a research environment, it also has potential for upscaling by employing drop-on-demand dispensing of polymer waveguide material in pre-patterned microchannels, for example in a roll-to-roll environment.

Severe Accident Test Station Activity Report

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

Pint, Bruce A.; Terrani, Kurt A.

2015-06-01

Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accidentmore » Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.« less

Holmium Doped Solid State Laser Resonantly Pumped and Q-Switched by Novel GaSb-Based Photonic Devices

DTIC Science & Technology

2011-08-31

dominant role of inter valence band absorption [7]. Details of the conduction band structure of the particular 0 20 40 60 80 100 0 10 20 30 CW 30s...here the n-cladding composition resulted into material with three valleys in conduction band to have almost the same energy minimum so no inter...emitting GaSb -based diode lasers was improved by utilization of the waveguide structure with asymmetric claddings. The AlGaAsSb p-cladding contained

SiC/SiC Cladding Materials Properties Handbook

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

Snead, Mary A.; Katoh, Yutai; Koyanagi, Takaaki

When a new class of material is considered for a nuclear core structure, the in-pile performance is usually assessed based on multi-physics modeling in coordination with experiments. This report aims to provide data for the mechanical and physical properties and environmental resistance of silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites for use in modeling for their application as accidenttolerant fuel cladding for light water reactors (LWRs). The properties are specific for tube geometry, although many properties can be predicted from planar specimen data. This report presents various properties, including mechanical properties, thermal properties, chemical stability under normal and offnormalmore » operation conditions, hermeticity, and irradiation resistance. Table S.1 summarizes those properties mainly for nuclear-grade SiC/SiC composites fabricated via chemical vapor infiltration (CVI). While most of the important properties are available, this work found that data for the in-pile hydrothermal corrosion resistance of SiC materials and for thermal properties of tube materials are lacking for evaluation of SiC-based cladding for LWR applications.« less

Cracking of a layered medium on an elastic foundation under thermal shock

NASA Technical Reports Server (NTRS)

Rizk, Abd El-Fattah A.; Erdogan, Fazil

1988-01-01

The cladded pressure vessel under thermal shock conditions which is simulated by using two simpler models was studied. The first model (Model 1) assumes that, if the crack size is very small compared to the vessel thickness, the problem can be treated as a semi-infinite elastic medium bonded to a very thin layer of different material. However, if the crack size is of the same order as the vessel thickness, the curvature effects may not be negligible. In this case it is assumed that the relatively thin walled hollow cylinder with cladding can be treated as a composite beam on an elastic foundation (Model 2). In both models, the effect of surface cooling rate is studied by assuming the temperature boundary condition to be a ramp function. The calculated results include the transient temperature, thermal stresses in the uncracked medium and stress intensity factors which are presented as a function of time, and the duration of cooling ramp. The stress intensity factors are also presented as a function of the size and the location of the crack. The problem is solved for two bonded materials of different thermal and mechanical properties. The mathematical formulation results in two singular integral equations which are solved numerically. The results are given for two material pairs, namely an austenitic steel layer welded on a ferritic steel substrate, and a ceramic coating on ferritic steel. In the case of the yielded clad, the stress intensity factors for a crack under the clad are determined by using a plastic strip model and are compared with elastic clad results.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

NASA Astrophysics Data System (ADS)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng; Alankar, Alankar; Subramanian, Gopinath; Stanek, Christopher

2017-01-01

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

Seamless metal-clad fiber-reinforced organic matrix composite structures and process for their manufacture

NASA Technical Reports Server (NTRS)

Bluck, Raymond M. (Inventor); Bush, Harold G. (Inventor); Johnson, Robert R. (Inventor)

1990-01-01

A metallic outer sleeve is provided which is capable of enveloping a hollow metallic inner member having continuous reinforcing fibers attached to the distal end thereof. The inner member is then introduced into outer sleeve until inner member is completely enveloped by outer sleeve. A liquid matrix member is then injected into space between inner member and outer sleeve. A pressurized heat transfer medium is flowed through the inside of inner member, thereby forming a fiber reinforced matrix composite material. The wall thicknesses of both inner member and outer sleeve are then reduced to the appropriate size by chemical etching, to adjust the thermal expansion coefficient of the metal-clad composite structure to the desired value. thereby forming a fiber reinforced matrix composite material. The wall thicknesses of both inner member and outer sleeve are then reduced to the appropriate size by chemical etching, to adjust the thermal expansion coefficient of the metal-clad composite structure to the desired value. The novelty of this invention resides in the development of a efficient method of producing seamless metal clad fiber reinforced organic matrix composite structures.

The effect of rider weight and additional weight in Icelandic horses in tölt: part II. Stride parameters responses.

PubMed

Gunnarsson, V; Stefánsdóttir, G J; Jansson, A; Roepstorff, L

2017-09-01

This study investigated the effects of rider weight in the BW ratio (BWR) range common for Icelandic horses (20% to 35%), on stride parameters in tölt in Icelandic horses. The kinematics of eight experienced Icelandic school horses were measured during an incremental exercise test using a high-speed camera (300 frames/s). Each horse performed five phases (642 m each) in tölt at a BWR between rider (including saddle) and horse starting at 20% (BWR20) and increasing to 25% (BWR25), 30% (BWR30), 35% (BWR35) and finally 20% (BWR20b) was repeated. One professional rider rode all horses and weight (lead) was added to saddle and rider as needed. For each phase, eight strides at speed of 5.5 m/s were analyzed for stride duration, stride frequency, stride length, duty factor (DF), lateral advanced placement, lateral advanced liftoff, unipedal support (UPS), bipedal support (BPS) and height of front leg action. Stride length became shorter (Y=2.73-0.004x; P0.05). In conclusion, increased BWR decreased stride length and increased DF proportionally to the same extent in all limbs, whereas BPS increased at the expense of decreased UPS. These changes can be expected to decrease tölt quality when subjectively evaluated according to the breeding goals for the Icelandic horse. However, beat, symmetry and height of front leg lifting were not affected by BWR.

Influence of Laser Power on the Shape of Single Tracks in Scanner Based Laser Wire Cladding

NASA Astrophysics Data System (ADS)

Barroi, A.; Gonçalves, D. Albertazzi; Hermsdorf, J.; Kaierle, S.; Overmeyer, L.

The shape of the cladding tracks is extremely important for producing layers or structures by adding them sequently. This paper shows the influence of the laser power of a diode laser in the range of 500 to 1000 W on the shapes of single tracks in scanner based laser wire cladding. The scanner was used to oscillate the beam perpendiculary to the welding direction. Stainless steel (ER 318 Si) wire with a 0.6 mm diameter was used as deposition material. Height, width, penetration, molten area and weld seam angles of single tracks were obtained from cross-sections at three different positions of each track. The influence of these different positions on the results depends on the traverse speed. The paper discusses this influence in respect to the heat dissipation in the substrate material.

Recovering tubewise power from three-dimensional nodal kinetics calculation during material relocation in an HWR

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

Kalimullah; Morris, E.E.; Yang, W.S.

1994-12-31

To analyze severe accidents in some special-purpose heavy-water reactors made of assemblies consisting of a number of coaxial tubes of aluminum-clad U-Al fuel and aluminum-clad neutron-capturing material, a mechanistic model, MARTINS, for tube beatup, melting, and molten material relocation has been developed and integrated with the DIF3D nodal hexagonal-z reactor kinetics and other phenomenological modules. The DIF3D kinetics homogenizes all materials located and computes the total power produced in an axial segment of a fuel assembly. This paper presents an approximate method, used in MARTINS, to calculate the distribution of this total nodal power into the intact fuel and capturingmore » material tubes and the meat-cladding mixtures relocating during tube disruption. The method accounts for the change in intraassembly radial power profile due to assembly geometry change with the progress of segment-by-segment disruption of different tubes. Earlier methods to recover pinwise power from nodal calculation for liquid-metal-cooled reactors and light water reactors (X-Y and hexagonal unit cells) are not practical for a disrupting assembly having material relocation. Figure 1 shows the assembly`s end view, divided into rings for modeling and analysis. A ring is a coolant subchannel plus the outer surrounding tube. The present method for distributing the nodal power consists of two parts: (a) calculation of the relative values of ring-by-ring power per unit uranium mass and power per unit mass of neutron-capturing material in a given assembly segment, and (b) normalization of these relative values such that the total power of all rings (intact tubes and U-Al-Cp meat-cladding mixtures, where Cp implies the neutron-capturing material) equals the DIF3D-calculated nodal power for the assembly axial segment.« less

Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions

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

Yang, Yong; Phillpot, Simon

Fuel cladding chemical interactions (FCCI) have been acknowledged as a critical issue in a metallic fuel/steel cladding system due to the formation of low melting intermetallic eutectic compounds between the fuel and cladding steel, resulting in reduction in cladding wall thickness as well as a formation of eutectic compounds that can initiate melting in the fuel at lower temperature. In order to mitigate FCCI, diffusion barrier coatings on the cladding inner surface have been considered. In order to generate the required coating techniques, pack cementation, electroplating, and electrophoretic deposition have been investigated. However, these methods require a high processing temperaturemore » of above 700 oC, resulting in decarburization and decomposition of the martensites in a ferritic/martensitic (F/M) cladding steel. Alternatively, organometallic chemical vapor deposition (OMCVD) can be a promising process due to its low processing temperature of below 600 oC. The aim of the project is to conduct applied and fundamental research towards the development of diffusion barrier coatings on the inner surface of F/M fuel cladding tubes. Advanced cladding steels such as T91, HT9 and NF616 have been developed and extensively studied as advanced cladding materials due to their excellent irradiation and corrosion resistance. However, the FCCI accelerated by the elevated temperature and high neutron exposure anticipated in fast reactors, can have severe detrimental effects on the cladding steels through the diffusion of Fe into fuel and lanthanides towards into the claddings. To test the functionality of developed coating layer, the diffusion couple experiments were focused on using T91 as cladding and Ce as a surrogate lanthanum fission product. By using the customized OMCVD coating equipment, thin and compact layers with a few micron between 1.5 µm and 8 µm thick and average grain size of 200 nm and 5 µm were successfully obtained at the specimen coated between 300oC and 500 oC, respectively. The coating layer contains both carbon and vanadium elements as quantified by WED, and the phases mainly consist of a mixture of V2C and VC, which was confirmed using X-ray diffraction patterns. In addition, the ratio between V and C varies with processing temperature, and it was observed that a higher temperature promotes the carbon adsorption and increases thickness of the coating. With optimized deposition conditions, we can apply the coating technique toward the actual T91 cladding materials, and provide the possibilities for the real application in sodium-cooled fast reactors (SFRs). Diffusion couple experiments were performed at both 550 oC and 690 oC, which corresponds to normal and aggressive operating temperatures, respectively. The results show that vanadium carbide coating with wider thickness (8 µm) and lower carbon concentration (27 at.%) reduced the width of the inter diffusion region, indicating that vanadium carbide coating can mitigate FCCI effectively. In specific, inter-diffusion between Fe and Ce was prohibited over most area, but Ce diffusion occurred toward the coating and the Fe substrate through thinner coating layer, which needs further optimization in terms of uniform coating thickness. Overall, it is concluded that this coating process can be successfully applied onto the inner surface of HT9 cladding tubes and the FCCI can be effectively mitigated if not totally eliminated.« less

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

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

Durbin, Samuel G.; Lindgren, Eric Richard

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing themore » internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. The arrangement of ducting was used to mimic conditions for an aboveground storage configuration in a vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured for a wide range of decay power and helium cask pressures. Of particular interest was the evaluation of the effect of increased helium pressure on peak cladding temperatures (PCTs) for identical thermal loads. All steady state peak temperatures and induced flow rates increased with increasing assembly power. Peak cladding temperatures decreased with increasing internal helium pressure for a given assembly power, indicating increased internal convection. In addition, the location of the PCT moved from near the top of the assembly to ~1/3 the height of the assembly for the highest (8 bar absolute) to the lowest (0 bar absolute) pressure studied, respectively. This shift in PCT location is consistent with the varying contribution of convective heat transfer proportional with of internal helium pressure.« less

SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

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

IJ van Rooyen; WR Lloyd; TL Trowbridge

2013-09-01

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designsmore » being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.« less

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

DOE PAGES

Cinbiz, M. Nedim; Brown, Nicholas R.; Terrani, Kurt A.; ...

2017-06-03

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failuremore » greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of integral tests for current and potentially accident tolerant cladding materials.« less

Interfacial Characterization of Dissimilar Joints Between Al/Mg/Al-Trilayered Clad Sheet to High-Strength Low-Alloy Steel

NASA Astrophysics Data System (ADS)

Macwan, A.; Jiang, X. Q.; Chen, D. L.

2015-07-01

Magnesium (Mg) alloys are increasingly used in the automotive and aerospace sectors to reduce vehicle weight. Al/Mg/Al tri-layered clad sheets are deemed as a promising alternative to improve the corrosion resistance and formability of Mg alloys. The structural application of Al/Mg/Al tri-layered clad sheets inevitably involves welding and joining in the multi-material vehicle body manufacturing. This study aimed to characterize the bonding interface microstructure of the Al/Mg/Al-clad sheet to high-strength low-alloy steel with and without Zn coating using ultrasonic spot welding at different levels of welding energy. It was observed that the presence of Zn coating improved the bonding at the interface due to the formation of Al-Zn eutectic structure via enhanced diffusion. At a higher level of welding energy, characteristic flow patterns of Zn into Al-clad layer were observed with an extensive penetration mainly along some high angle grain boundaries. The dissimilar joints without Zn coating made at a high welding energy of 800 J failed partially from the Al/Fe weld interface and partially from the Al/Mg clad interface, while the joints with Zn coating failed from the Al/Mg clad interface due to the presence of brittle Al12Mg17 phase.

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

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

Cinbiz, M. Nedim; Brown, Nicholas R.; Terrani, Kurt A.

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failuremore » greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of integral tests for current and potentially accident tolerant cladding materials.« less

Laser cladding assisted by friction stir processing for preparation of deformed crack-free Ni-Cr-Fe coating with nanostructure

NASA Astrophysics Data System (ADS)

Xie, Siyao; Li, Ruidi; Yuan, Tiechui; Chen, Chao; Zhou, Kechao; Song, Bo; Shi, Yusheng

2018-02-01

Although laser cladding has find its widespread application in surface hardening, this technology has been significantly limited by the solidification crack, which usually initiates along grain boundary due to the brittle precipitation in grain boundary and networks formation during the laser rapid melting/solidification process. This paper proposed a novel laser cladding technology assisted by friction stir processing (FSP) to eliminate the usual metallurgical defects by the thermomechanical coupling effect of FSP with the Ni-Cr-Fe as representative coating material. By the FSP assisted laser cladding, the crack in laser cladding Ni-Cr-Fe coating was eliminated and the coarse networks of laser cladding coating was transformed into dispersed nanoparticles. Moreover, the plastic layers with thicknesses 47-140 μm can be observed, with gradient grain refinement from substrate to the top surface in which grain size reached 300 nm and laser photocoagulation net second phase crushed in the layer. In addition, cracks closed in the plastic zone. The refinement of grain resulted the hardness increased to over 400 HV, much higher than the 300 HV of the laser cladding structure. After FSP, the friction coefficient decreased from 0.6167 to 0.5645 which promoted the wear resistance.

Shielding gas effect to diffusion activities of magnesium and copper on aluminum clad

NASA Astrophysics Data System (ADS)

Manurung, Charles SP; Napitupulu, Richard AM

2017-09-01

Aluminum is the second most metal used in many application, because of its corrosion resistance. The Aluminum will be damaged in over time if it’s not maintained in good condition. That is important to give protection to the Aluminums surface. Cladding process is one of surface protection methodes, especially for metals. Aluminum clad copper (Al/Cu) or copper clad aluminum (Cu/Al) composite metals have been widely used for many years. These mature protection method and well tested clad metal systems are used industrially in a variety application. The inherent properties and behavior of both copper and aluminum combine to provide unique performance advantages. In this paper Aluminum 2024 series will be covered with Aluminum 1100 series by hot rolling process. Observations will focus on diffusion activities of Mg and Cu that not present on Aluminum 1100 series. The differences of clad material samples is the use of shielding gas during heating before hot rolling process. The metallurgical characteristics will be examined by using optical microscopy. Transition zone from the interface cannot be observed but from Energy Dispersive Spectrometry it’s found that Mg and Cu are diffused from base metal (Al 2024) to the clad metal (Al 1100). Hardness test proved that base metals hardness to interface was decrease.

Development and Validation of Accident Models for FeCrAl Cladding

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

Gamble, Kyle Allan Lawrence; Hales, Jason Dean

2016-08-01

The purpose of this milestone report is to present the work completed in regards to material model development for FeCrAl cladding and highlight the results of applying these models to Loss of Coolant Accidents (LOCA) and Station Blackouts (SBO). With the limited experimental data available (essentially only the data used to create the models) true validation is not possible. In the absence of another alternative, qualitative comparisons during postulated accident scenarios between FeCrAl and Zircaloy-4 cladded rods have been completed demonstrating the superior performance of FeCrAl.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

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

Montgomery, Robert, E-mail: robert.montgomery@pnnl.gov; Tomé, Carlos, E-mail: tome@lanl.gov; Liu, Wenfeng, E-mail: wenfeng.liu@anatech.com

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC)more » polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.« less

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

Surface protection of light metals by one-step laser cladding with oxide ceramics

NASA Astrophysics Data System (ADS)

Nowotny, S.; Richter, A.; Tangermann, K.

1999-06-01

Today, intricate problems of surface treatment can be solved through precision cladding using advanced laser technology. Metallic and carbide coatings have been produced with high-power lasers for years, and current investigations show that laser cladding is also a promising technique for the production of dense and precisely localized ceramic layers. In the present work, powders based on Al2O3 and ZrO2 were used to clad aluminum and titanium light alloys. The compact layers are up to 1 mm thick and show a nonporous cast structure as well as a homogeneous network of vertical cracks. The high adhesive strength is due to several chemical and mechanical bonding mechanisms and can exceed that of plasmasprayed coatings. Compared to thermal spray techniques, the material deposition is strictly focused onto small functional areas of the workpiece. Thus, being a precision technique, laser cladding is not recommended for large-area coatings. Examples of applications are turbine components and filigree parts of pump casings.

10J water-cooled DPSSL system based on Yb:YAG crystal edge-cladded by Cr:YAG ceramics (Conference Presentation)

NASA Astrophysics Data System (ADS)

Zheng, Jian-Gang; Yan, Xiongwei; Jiang, Xinying; Wang, Zhenguo; Li, Mingzhong; Zhang, Jun; Zhu, Qihua; Zheng, Wanguo

2017-05-01

Laser Inertial Fusion Energy (IFE) has been attracting the interests of the researchers around the world, because of the promising to the future energy. The Yb:YAG was broadly used in the research field of high-peak power and large energy laser with repetition-rate for IFE because of its outstanding performance, including significant thermal and mechanical capacities, long upper energy level lifetime, high quantum efficiency and highly doping capacity. But it exhibits high saturation fluence at room temperature because of the small emission and absorption cross-section. And at the same time this gain material exhibits self-absorption of laser because of the thermal population at lower laser level at room temperature. Ant it appears to have been solved by means of the cryogenic temperature, but the total efficiency of the laser system will be decreased as the use of cryogenic temperature. The amplified spontaneous emission (ASE) effect of the amplifier can be relaxed by means of edge-cladded absorption material. And the difficulties of edge cladding can be will solved as the emergence of ceramics. But at present the ceramics exhibits high scattering and many disfigurements, which limited the application in the high-power large-energy laser system. So the edge-cladding of Yb:YAG crystal will be a key issue for solution the ASE in amplifier. In this paper, we will introduce a 10J water-cooled DPSSL system, based on Yb:YAG crystal at room temperature. In this system a new edge cladding method has been used, that the Yb:YAG crystal was edge cladded by Cr:YAG ceramics, which was used as the absorption material of ASE. The amplifier was an active mirror water-cooled room temperature amplifier. With the help of this edge cladding the ASE has been lowered, and about 5 times small signal gain has been obtained in a single pass amplification, which was much higher than the earlier of 2 times. And the wavefront aberrance of the laser beam was also reduced due to the thermal equilibrium between the edge cladding and the gain region. the amplifiers can be stably operated under 10Hz. Finally the output of the laser system was about 7.15J@10Hz and 10.8J@1-2Hz. The total optical-to-optical efficiency was about 8.3% for 1-2Hz (under the condition of 120kW/1ms pumping, 880mJ input and 10.8J output) and 5.6% for 10Hz.

Development of a reliable estimation procedure of radioactivity inventory in a BWR plant due to neutron irradiation for decommissioning

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi; Ueno, Jun

2017-09-01

Reliable information of radioactivity inventory resulted from the radiological characterization is important in order to plan decommissioning planning and is also crucial in order to promote decommissioning in effectiveness and in safe. The information is referred to by planning of decommissioning strategy and by an application to regulator. Reliable information of radioactivity inventory can be used to optimize the decommissioning processes. In order to perform the radiological characterization reliably, we improved a procedure of an evaluation of neutron-activated materials for a Boiling Water Reactor (BWR). Neutron-activated materials are calculated with calculation codes and their validity should be verified with measurements. The evaluation of neutron-activated materials can be divided into two processes. One is a distribution calculation of neutron-flux. Another is an activation calculation of materials. The distribution calculation of neutron-flux is performed with neutron transport calculation codes with appropriate cross section library to simulate neutron transport phenomena well. Using the distribution of neutron-flux, we perform distribution calculations of radioactivity concentration. We also estimate a time dependent distribution of radioactivity classification and a radioactive-waste classification. The information obtained from the evaluation is utilized by other tasks in the preparatory tasks to make the decommissioning plan and the activity safe and rational.

In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR

NASA Astrophysics Data System (ADS)

Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka

The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

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

Trianti, Nuri, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Su'ud, Zaki, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Arif, Idam, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id

2014-09-30

Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tightmore » concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.« less

Comments on ""Contact Diffusion Interaction of Materials with Cladding''

NASA Technical Reports Server (NTRS)

Morris, J. F.

1972-01-01

A Russian paper by A. A. Babad-Zakhryapina contributes much to the understanding of fuel, clad interactions, and thus to nuclear thermionic technology. In that publication the basic diffusion expression is a simple one. A more general but complicated equation for this mass transport results from the present work. With appropriate assumptions, however, the new relation reduces to Babad-Zakhryapina's version.

Thermal load histories for North American roof assembles using various cladding materials including wood-thermoplastic composite shingles

Treesearch

J. E. Winandy

2006-01-01

Since 1991, thermal load histories for various roof cladding types have been monitored in outdoor attic structures that simulate classic North American light-framed construction. In this paper, the 2005 thermal loads for wood-based composite roof sheathing, wood rafters, and attics under wood-plastic composite shingles are compared to common North American roof...

Status of Wrought FeCrAl-UO 2 Capsules Irradiated in the Advanced Test Reactor

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

Field, Kevin G.; Harp, J.; Core, G.

2017-07-01

Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction,more » and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.« less

Bioactivity of fluorapatite/alumina composite coatings deposited on Ti6Al4V substrates by laser cladding

NASA Astrophysics Data System (ADS)

Chien, C. S.; Liu, C. W.; Kuo, T. Y.; Wu, C. C.; Hong, T. F.

2016-04-01

Hydroxyapatite (HA) is one of the most commonly used coating materials for metal implants. However, following high-temperature deposition, HA easily decomposes into an unstable phase or forms an amorphous phase, and hence, the long-term stability of the implant is reduced. Accordingly, the present study investigates the use of fluorapatite (FA) fortified with 20 wt% alumina (α-Al2O3) as an alternative biomedical coating material. The coatings are deposited on Ti6Al4V substrates using a Nd:YAG laser cladding process performed with laser powers and travel speeds of 400 W/200 mm/min, 800 W/400 mm/min and 1200 W/600 mm/min, respectively. The results show that for all of the specimens, a strong metallurgical bond is formed at the interface between the coating layer and the transition layer due to melting and diffusion. The XRD analysis results reveal that the cladding layers in all of the specimens consist mainly of FA, β-TCP, CaF2, Ti and θ-Al2O3 phases. In addition, the cladding layers of the specimens prepared using laser powers of 400 and 800 W also contain CaTiO3 and CaAl2O4, while that of the specimen clad using a power of 1200 W contains TTCP and CaO. Following immersion in simulated body fluid for 14 days, all of the specimens precipitate dense bone-like apatite and exhibit excellent bioactivity. However, among all of the specimens, the specimen that is prepared with a laser power of 800 W shows the best biological activity due to the presence of residual FA, apatite-generating CaTiO3 and a rough cladding layer surface.

Microstructure and corrosion resistance of TC2 Ti alloy by laser cladding with Ti/TiC/TiB2 powders

NASA Astrophysics Data System (ADS)

Diao, Yunhua; Zhang, Kemin

2015-10-01

In the present work, a TiC/TiB2 composite coating was produced onto a TC2 Ti alloy by laser cladding with Ti/TiC/TiB2 powders. The surface microstructure, phase components and compositions were characterized with methods of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffractometry (XRD), and energy dispersive spectrometry (EDS). The cladding layer is consisted of Ti, TiC and TiB2. And the surface microhardness was measured. After laser cladding, a maximum hardness of 1100 HV is achieved in the laser cladding surface layer, which is more three times higher than that of the TC2 substrate (∼300 HV). Due to the formation of TiC and TiB2 intermetallic compounds in the alloyed region and grain refinement, the microhardness of coating is higher than TC2 Ti alloy. In this paper, the corrosion property of matrix material and treated samples were both measured in NaCl (3.5 wt%) aqueous solution. From the result we can see that the laser cladding specimens' corrosion property is clearly becoming better than that of the substrate.

Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

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

Siefken, L.J.

1999-01-01

Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocation of melted cladding in the circumferential direction. Then, rules based on PIE results were established for calculating the effect of cladding that has relocated from abovemore » on the oxidation and integrity of the lower intact cladding upon which it solidifies. Next, three different methods were identified for calculating the extent of dissolution of the oxidic part of the cladding due to its contact with the metallic part. The extent of dissolution effects the stress and thus the integrity of the oxidic part of the cladding. Then, an empirical equation was presented for calculating the stress in the oxidic part of the cladding and evaluating its integrity based on this calculated stress. This empirical equation replaces the current criterion for loss of integrity which is based on temperature and extent of oxidation. Finally, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The implementation of these models is expected to eliminate the tendency of the SCDAP/RELAP5 code to overpredict the extent of oxidation of the upper part of fuel rods and to underpredict the extent of oxidation of the lower part of fuel rods and the part with a high concentration of relocated material. This report is a revision and reissue of the report entitled, Improvements in Modeling of Cladding Oxidation and Meltdown.« less

Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

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

Zhang, Yongfeng

2016-09-01

U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significantmore » progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.« less

Structure and Thermodynamical Properties of Zirconium Hydrides from First-Principle

NASA Astrophysics Data System (ADS)

Blomqvist, Jakob; Olofsson, Johan; Alvarez, Anna-Maria; Bjerkén, Christina

Zirconium alloys are used as nuclear fuel cladding material due to their mechanical and corrosion resistant properties together with their favorable cross-section for neutron scattering. At running conditions, however, there will be an increase of hydrogen in the vicinity of the cladding surface at the water side of the fuel. The hydrogen will diffuse into the cladding material and at certain conditions, such as lower temperatures and external load, hydrides will precipitate out in the material and cause well known embrittlement, blistering and other unwanted effects. Using phase-field methods it is now possible to model precipitation buildup in metals, for example as a function of hydrogen concentration, temperature and external load, but the technique relies on input of parameters, such as the formation energy of the hydrides and matrix. To that end, we have computed, using the density functional theory (DFT) code GPAW, the latent heat of fusion as well as solved the crystal structure for three zirconium hydride polymorphs: δ-ZrH1.6, γ-ZrH, and Є-ZrH2.

Test Plan for the Boiling Water Reactor Dry Cask Simulator

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

Durbin, Samuel; Lindgren, Eric R.

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing themore » internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below-ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.« less

Accident-tolerant oxide fuel and cladding

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

Mariani, Robert D.

Systems and methods for accident tolerant oxide fuel. One or more disks can be placed between fuel pellets comprising UO.sub.2, wherein such disks possess a higher thermal conductivity material than that of the UO.sub.2 to provide enhanced heat rejection thereof. Additionally, a cladding coating comprising zircaloy coated with a material that provides stability and high melting capability can be provided. The pellets can be configured as annular pellets having an annulus filled with the higher thermal conductivity material. The material coating the zircaloy can be, for example, Zr.sub.5Si.sub.4 or another silicide such as, for example, a Zr-Silicide that limits corrosion.more » The aforementioned higher thermal conductivity material can be, for example, Si, Zr.sub.xSi.sub.y, Zr, or Al.sub.2O.sub.3.« less

Compatibility of refractory materials for nuclear reactor poison control systems

NASA Technical Reports Server (NTRS)

Sinclair, J. H.

1974-01-01

Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

Recent developments in BWR fuel design

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

Congdon, S.P.; Noble, L.D.; Wood, J.E.

1991-11-01

Substantial increases in the cost effectiveness and performance capability of boiling water reactor (BWR) fuel designs have been implemented in the past 5 to 7 yr. This increase has been driven by (a) utility desires to lower fuel and operating costs and (b) design innovations that have lowered enrichment requirements, improved thermal-hydraulic performance, and increased discharge exposure. Higher discharge exposures reduce disposal costs for European and Asian utilities and enable US utilities to lengthen operating cycles. A typical BWR reload fuel bundle fabricated today has 25% higher {sup 235}U enrichment and a factor of 2 higher gadolinium loading than onemore » made several years ago. Today's BWR fuel bundles also contain more unheated water reduces the axial water density variation, lowers the void coefficient, and enhances the neutron efficiency of the bundle, reducing both the gadolinium poison and the enrichment requirements. In addition to these general trends, the following unique design innovations have further enhanced the fuel cost efficiency and performance characteristics of BWR fuel: ferrule spacer, part length rods, interactive channel, and bundle enhanced spectral shift. GE's fuel designs offer the flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility and fuel cycle economics.« less

Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

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

Arai, Kenji; Ebata, Shigeo

1997-07-01

This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding ofmore » the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.« less

The effect of rider weight and additional weight in Icelandic horses in tölt: part I. Physiological responses.

PubMed

Stefánsdóttir, G J; Gunnarsson, V; Roepstorff, L; Ragnarsson, S; Jansson, A

2017-09-01

This study examined the effect of increasing BW ratio (BWR) between rider and horse, in the BWR range common for Icelandic horses (20% to 35%), on heart rate (HR), plasma lactate concentration (Lac), BWR at Lac 4 mmol/l (W4), breathing frequency (BF), rectal temperature (RT) and hematocrit (Hct) in Icelandic horses. In total, eight experienced school-horses were used in an incremental exercise test performed outdoors on an oval riding track and one rider rode all horses. The exercise test consisted of five phases (each 642 m) in tölt, a four-beat symmetrical gait, at a speed of 5.4±0.1 m/s (mean±SD), where BWR between rider (including saddle) and horse started at 20% (BWR20), was increased to 25% (BWR25), 30% (BWR30), and 35% (BWR35) and finally decreased to 20% (BWR20b). Between phases, the horses were stopped (~5.5 min) to add lead weights to specially adjusted saddle bags and a vest on the rider. Heart rate was measured during warm-up, the exercise test and after 5, 15 and 30 min of recovery and blood samples were taken and BF recorded at rest, and at end of each of these aforementioned occasions. Rectal temperature was measured at rest, at end of the exercise test and after a 30-min recovery period. Body size and body condition score (BCS) were registered and a clinical examination performed on the day before the test and for 2 days after. Heart rate and BF increased linearly (P0.05), but negative correlations (P<0.05) existed between body size measurements and Hct. While HR, Hct and BF recovered to values at rest within 30 min, Lac and RT did not. All horses had no clinical remarks on palpation and at walk 1 and 2 days after the test. In conclusion, increasing BWR from 20% to 35% resulted in increased HR, Lac, RT and BF responses in the test group of experienced adult Icelandic riding horses. The horses mainly worked aerobically until BWR reached 22.7%, but considerable individual differences (17.0% to 27.5%) existed that were not linked to horse size, but to back BCS.

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

NASA Astrophysics Data System (ADS)

Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

2014-06-01

The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Absorptivity Measurements and Heat Source Modeling to Simulate Laser Cladding

NASA Astrophysics Data System (ADS)

Wirth, Florian; Eisenbarth, Daniel; Wegener, Konrad

The laser cladding process gains importance, as it does not only allow the application of surface coatings, but also additive manufacturing of three-dimensional parts. In both cases, process simulation can contribute to process optimization. Heat source modeling is one of the main issues for an accurate model and simulation of the laser cladding process. While the laser beam intensity distribution is readily known, the other two main effects on the process' heat input are non-trivial. Namely the measurement of the absorptivity of the applied materials as well as the powder attenuation. Therefore, calorimetry measurements were carried out. The measurement method and the measurement results for laser cladding of Stellite 6 on structural steel S 235 and for the processing of Inconel 625 are presented both using a CO2 laser as well as a high power diode laser (HPDL). Additionally, a heat source model is deduced.

Restoration of Worn Movable Bridge Props with Use of Bronze Claddings.

PubMed

Viňáš, Ján; Vrabeľ, Marek; Greš, Miroslav; Brezina, Jakub; Sabadka, Dušan; Fedorko, Gabriel; Molnár, Vieroslav

2018-03-21

This article examined the possibility of using CuSn6P claddings in sliding bearing renovation of movable pontoon bridge props. The bronze layer was welded on cylinders of the high-strength steel S355J0WP EN 10155-93, in an inert atmosphere using an automated welding method (gas tungsten arc welding). Pulsed arc welding was used to minimize the effects of heat on the cladding area, while also accounting for the differences in the physical properties of the joined metals. The sliding bearing was created in two layers. The quality of the cladding layer was evaluated by nondestructive and/or destructive tests. The quality of the surface was assessed by visual inspection (visual testing) in accordance with the EN ISO 17637 standard. The quality of the claddings was evaluated by metallographic analysis, performed using light microscopy. The microhardness values of a few weld areas were determined by Vickers tests, performed according to the EN ISO 9015-2 standard. The analyses confirmed that the welding parameters and filler material used resulted in high-quality weld joints with no internal (subsurface) or metallurgical defects.

Restoration of Worn Movable Bridge Props with Use of Bronze Claddings

PubMed Central

Viňáš, Ján; Vrabeľ, Marek; Greš, Miroslav; Brezina, Jakub; Sabadka, Dušan; Fedorko, Gabriel

2018-01-01

This article examined the possibility of using CuSn6P claddings in sliding bearing renovation of movable pontoon bridge props. The bronze layer was welded on cylinders of the high-strength steel S355J0WP EN 10155-93, in an inert atmosphere using an automated welding method (gas tungsten arc welding). Pulsed arc welding was used to minimize the effects of heat on the cladding area, while also accounting for the differences in the physical properties of the joined metals. The sliding bearing was created in two layers. The quality of the cladding layer was evaluated by nondestructive and/or destructive tests. The quality of the surface was assessed by visual inspection (visual testing) in accordance with the EN ISO 17637 standard. The quality of the claddings was evaluated by metallographic analysis, performed using light microscopy. The microhardness values of a few weld areas were determined by Vickers tests, performed according to the EN ISO 9015–2 standard. The analyses confirmed that the welding parameters and filler material used resulted in high-quality weld joints with no internal (subsurface) or metallurgical defects. PMID:29561762

3D analysis of thermal and stress evolution during laser cladding of bioactive glass coatings.

PubMed

Krzyzanowski, Michal; Bajda, Szymon; Liu, Yijun; Triantaphyllou, Andrew; Mark Rainforth, W; Glendenning, Malcolm

2016-06-01

Thermal and strain-stress transient fields during laser cladding of bioactive glass coatings on the Ti6Al4V alloy basement were numerically calculated and analysed. Conditions leading to micro-cracking susceptibility of the coating have been investigated using the finite element based modelling supported by experimental results of microscopic investigation of the sample coatings. Consecutive temperature and stress peaks are developed within the cladded material as a result of the laser beam moving along the complex trajectory, which can lead to micro-cracking. The preheated to 500°C base plate allowed for decrease of the laser power and lowering of the cooling speed between the consecutive temperature peaks contributing in such way to achievement of lower cracking susceptibility. The cooling rate during cladding of the second and the third layer was lower than during cladding of the first one, in such way, contributing towards improvement of cracking resistance of the subsequent layers due to progressive accumulation of heat over the process. Copyright © 2016 Elsevier Ltd. All rights reserved.

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

NASA Astrophysics Data System (ADS)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

Parametric Study and Multi-Criteria Optimization in Laser Cladding by a High Power Direct Diode Laser

NASA Astrophysics Data System (ADS)

Farahmand, Parisa; Kovacevic, Radovan

2014-12-01

In laser cladding, the performance of the deposited layers subjected to severe working conditions (e.g., wear and high temperature conditions) depends on the mechanical properties, the metallurgical bond to the substrate, and the percentage of dilution. The clad geometry and mechanical characteristics of the deposited layer are influenced greatly by the type of laser used as a heat source and process parameters used. Nowadays, the quality of fabricated coating by laser cladding and the efficiency of this process has improved thanks to the development of high-power diode lasers, with power up to 10 kW. In this study, the laser cladding by a high power direct diode laser (HPDDL) as a new heat source in laser cladding was investigated in detail. The high alloy tool steel material (AISI H13) as feedstock was deposited on mild steel (ASTM A36) by a HPDDL up to 8kW laser and with new design lateral feeding nozzle. The influences of the main process parameters (laser power, powder flow rate, and scanning speed) on the clad-bead geometry (specifically layer height and depth of the heat affected zone), and clad microhardness were studied. Multiple regression analysis was used to develop the analytical models for desired output properties according to input process parameters. The Analysis of Variance was applied to check the accuracy of the developed models. The response surface methodology (RSM) and desirability function were used for multi-criteria optimization of the cladding process. In order to investigate the effect of process parameters on the molten pool evolution, in-situ monitoring was utilized. Finally, the validation results for optimized process conditions show the predicted results were in a good agreement with measured values. The multi-criteria optimization makes it possible to acquire an efficient process for a combination of clad geometrical and mechanical characteristics control.

Interpretation of the results of the CORA-33 dry core BWR test

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

Ott, L.J.; Hagen, S.

All BWR degraded core experiments performed prior to CORA-33 were conducted under ``wet`` core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ``dry`` core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ``dry`` core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions ofmore » a ``dry`` BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ``dry`` core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed.« less

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

DOE PAGES

Ward, Andrew; Downar, Thomas J.; Xu, Y.; ...

2015-04-22

The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

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

T.L. Lotz

1997-02-15

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercialmore » spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.« less

High Fidelity BWR Fuel Simulations

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

Yoon, Su Jong

This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fractionmore » and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.« less

Design solutions for the solar cell interconnect fatigue fracture problem

NASA Technical Reports Server (NTRS)

Mon, G. R.; Ross, R. G., Jr.

1982-01-01

Mechanical fatigue of solar cell interconnects is a major failure mechanism in photovoltaic arrays. A comprehensive approach to the reliability design of interconnects, together with extensive design data for the fatigue properties of copper interconnects, has been published. This paper extends the previous work, developing failure prediction (fatigue) data for additional interconnect material choices, including aluminum and a variety of copper-Invar and copper-steel claddings. An improved global fatigue function is used to model the probability-of-failure statistics of each material as a function of level and number of cycles of applied strain. Life-cycle economic analyses are used to evaluate the relative merits of each material choce. The copper-Invar clad composites demonstrate superior performance over pure copper. Aluminum results are disappointing.

Procurement of a Large Area Mapping FTIR Microscope for Organic-Inorganic Interfacial Analysis in Biological Materials

DTIC Science & Technology

2015-12-31

biological composites. This includes the chemical mapping of the radular teeth of Cryptochiton stelleri (chiton), the crush resistant exoskeleton ...mapping of the radular teeth of Cryptochiton stelleri (chiton), the crush resistant exoskeleton from Phloeodes diabolicus (the Iron Clad beetle), and the... exoskeleton from Phloeodes diabolicus (the Iron Clad beetle), and the hard and impact resistant dactyl club from the stomatopod Odontodactylus scyllarus

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

DOE PAGES

Gamble, Kyle A.; Barani, Tommaso; Pizzocri, David; ...

2017-04-30

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterionmore » is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.« less

Refractory clad transient internal probe for magnetic field measurements in high temperature plasmas

NASA Astrophysics Data System (ADS)

Kim, Hyundae; Cellamare, Vincent; Jarboe, Thomas R.; Mattick, Arthur T.

2005-05-01

The transient internal probe (TIP) is a diagnostic for local internal field measurements in high temperature plasmas. A verdet material, which rotates the polarization angle of the laser light under magnetic fields, is launched into a plasma at about 1.8km/s. A linearly polarized Ar+ laser illuminates the probe in transit and the light retroreflected from the probe is analyzed to determine the local magnetic field profiles. The TIP has been used for magnetic field measurements on the helicity injected torus where electron temperature Te⩽80eV. In order to apply the TIP in higher temperature plasmas, refractory clad probes have been developed utilizing a sapphire tube, rear disc, and a MgO window on the front. The high melting points of these refractory materials should allow probe operation at plasma electron temperatures up to Te˜300eV. A retroreflecting probe has also been developed using "catseye" optics. The front window is replaced with a plano-convex MgO lens, and the back surface of the probe is aluminized. This approach reduces spurious polarization effects and provides refractory cladding for the probe entrance face. In-flight measurements of a static magnetic field demonstrate the ability of the clad probes to withstand gun-launch acceleration, and provide high accuracy measurements of magnetic field.

Behavior of polymer cladding materials under extremely high temperatures

NASA Astrophysics Data System (ADS)

Clark, Timothy E.; Chang, Selee; Kwak, SeungJo; Oh, Jung Hyun

2012-01-01

Polymer claddings with low refractive indices for silica core fibers were developed. Applications include fiber lasers and transmission of high power lasers in surgery. For many applications, operating fibers under high temperatures is desirable. In a previous publication, the results of testing polymer cladded silica core fiber at 150°C for 6400 hours were given, along with 5000 hours of testing polymer films. The results at 150°C were encouraging, with little additional loss measured. Here we test polymers under more severe conditions, at 270°C, for periods up to 10 hours. The polymers' cured indices range from 1.374 to 1.397 (at 852 nm). Changes in Young's modulus, refractive index, yellowing, weight, hardness, strength, and elongation were observed. While these polymers cannot function at 270°C for extended periods, it is possible to expose them for shorter durations without significant damage. Some polymer properties actually improved after 4 hours of heating. Fibers clad with such polymers have been successfully jacketed with extruded materials, and have endured high temperatures for a few minutes. It is possible that a sensor, fiber laser or other fiber device could function in these temperatures for short periods without the coating properties changing beyond values required for operation.

Using wire shaping techniques and holographic optics to optimize deposition characteristics in wire-based laser cladding.

PubMed

Goffin, N J; Higginson, R L; Tyrer, J R

2016-12-01

In laser cladding, the potential benefits of wire feeding are considerable. Typical problems with the use of powder, such as gas entrapment, sub-100% material density and low deposition rate are all avoided with the use of wire. However, the use of a powder-based source material is the industry standard, with wire-based deposition generally regarded as an academic curiosity. This is because, although wire-based methods have been shown to be capable of superior quality results, the wire-based process is more difficult to control. In this work, the potential for wire shaping techniques, combined with existing holographic optical element knowledge, is investigated in order to further improve the processing characteristics. Experiments with pre-placed wire showed the ability of shaped wire to provide uniformity of wire melting compared with standard round wire, giving reduced power density requirements and superior control of clad track dilution. When feeding with flat wire, the resulting clad tracks showed a greater level of quality consistency and became less sensitive to alterations in processing conditions. In addition, a 22% increase in deposition rate was achieved. Stacking of multiple layers demonstrated the ability to create fully dense, three-dimensional structures, with directional metallurgical grain growth and uniform chemical structure.

Using wire shaping techniques and holographic optics to optimize deposition characteristics in wire-based laser cladding

PubMed Central

Higginson, R. L.; Tyrer, J. R.

2016-01-01

In laser cladding, the potential benefits of wire feeding are considerable. Typical problems with the use of powder, such as gas entrapment, sub-100% material density and low deposition rate are all avoided with the use of wire. However, the use of a powder-based source material is the industry standard, with wire-based deposition generally regarded as an academic curiosity. This is because, although wire-based methods have been shown to be capable of superior quality results, the wire-based process is more difficult to control. In this work, the potential for wire shaping techniques, combined with existing holographic optical element knowledge, is investigated in order to further improve the processing characteristics. Experiments with pre-placed wire showed the ability of shaped wire to provide uniformity of wire melting compared with standard round wire, giving reduced power density requirements and superior control of clad track dilution. When feeding with flat wire, the resulting clad tracks showed a greater level of quality consistency and became less sensitive to alterations in processing conditions. In addition, a 22% increase in deposition rate was achieved. Stacking of multiple layers demonstrated the ability to create fully dense, three-dimensional structures, with directional metallurgical grain growth and uniform chemical structure. PMID:28119550

In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

DOE PAGES

Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; ...

2015-08-25

Here, one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filteringmore » unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.« less

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

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

Gamble, Kyle A.; Barani, Tommaso; Pizzocri, David

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterionmore » is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.« less

Outer skin protection of columbium Thermal Protection System (TPS) panels

NASA Technical Reports Server (NTRS)

Culp, J. D.

1973-01-01

A coated columbium alloy material system 0.04 centimeter thick was developed which provides for increased reliability to the load bearing character of the system in the event of physical damage to and loss of the exterior protective coating. The increased reliability to the load bearing columbium alloy (FS-85) was achieved by interposing an oxidation resistant columbium alloy (B-1) between the FS-85 alloy and a fused slurry silicide coating. The B-1 alloy was applied as a cladding to the FS-85 and the composite was fused slurry silicide coated. Results of material evaluation testing included cyclic oxidation testing of specimens with intentional coating defects, tensile testing of several material combinations exposed to reentry profile conditions, and emittance testing after cycling of up to 100 simulated reentries. The clad material, which was shown to provide greater reliability than unclad materials, holds significant promise for use in the thermal protection system of hypersonic reentry vehicles.

Process for the manufacture of seamless metal-clad fiber-reinforced organic matrix composite structures

NASA Technical Reports Server (NTRS)

Bluck, Raymond M. (Inventor); Bush, Harold G. (Inventor); Johnson, Robert R. (Inventor)

1991-01-01

A process for producing seamless metal-clad composite structures includes providing a hollow, metallic inner member and an outer sleeve to surround the inner member and define an inner space therebetween. A plurality of continuous reinforcing fibers is attached to the distal end of the outside diameter of the inner member, and the inner member is then introduced, distal end first, into one end of the outer sleeve. The inner member is then moved, distal end first, into the outer sleeve until the inner member is completely enveloped by the outer sleeve. A liquid matrix material is then injected into the space containing the reinforcing fibers between the inner member and the outer sleeve. Next a pressurized heat transfer medium is passed through the inner member to cure the liquid matrix material. Finally, the wall thickness of both the inner member and the outer sleeve are reduced to desired dimensions by chemical etching, which adjusts the thermal expansion coefficient of the metal-clad composite structure to a desired value.

RERTR-9 Summary Report

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

D. M. Perez

2011-05-01

The RERTR-9 experiment was designed to test the effect of modified fuel/clad interfaces in monolithic fuel plates and to demonstrate that the addition of Si to the matrix material in dispersion plates continued to be effective at high loading (~8.5 g U/cc). Several monolithic fuel plates were fabricated by Hot Isostatic Pressing (HIP) and Friction Bonding (FB) with thin layers of Si inserted and by HIP with a Zr diffusion barrier between the fuel and cladding. Si was applied to the interface by thermal spray of Al Si mixtures and by the insertion of thin Si-rich Al alloy foil betweenmore » the fuel/clad interface. The dispersion fuel plates were fabricated by semi-standard rolling techniques (the reduction by rolling was lowered to limit fabrication defects). Matrix materials consisted of Al-Si alloys and mixtures with various levels of Si. The following report summarizes the life of the RERTR-9A/B experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.« less

Suspended silica beam splitters on silicon with large core-clad index deference

NASA Astrophysics Data System (ADS)

Zhang, Xiaomin; Armani, Andrea M.

2012-03-01

Optical beam splitters form a fundamental component in integrated optical systems, performing as modulators, interferometers and (de)multiplexers. While silica is a desirable material, because of its low non-linear susceptibility, it is extremely challenging to achieve the requisite core-clad refractive index contrast. In this work, silica splitters with an effective refractive index difference of 25% between the core and clad is demonstrated. The splitter can divide power evenly with low crosstalk from 1520 to 1630nm. In addition, the splitting ratio doesn't change and the output power increases linearly with the input power, which indicates a low susceptibility to thermal effects. The splitter's polarization independent behavior is also verified.

Real-time laser cladding control with variable spot size

NASA Astrophysics Data System (ADS)

Arias, J. L.; Montealegre, M. A.; Vidal, F.; Rodríguez, J.; Mann, S.; Abels, P.; Motmans, F.

2014-03-01

Laser cladding processing has been used in different industries to improve the surface properties or to reconstruct damaged pieces. In order to cover areas considerably larger than the diameter of the laser beam, successive partially overlapping tracks are deposited. With no control over the process variables this conduces to an increase of the temperature, which could decrease mechanical properties of the laser cladded material. Commonly, the process is monitored and controlled by a PC using cameras, but this control suffers from a lack of speed caused by the image processing step. The aim of this work is to design and develop a FPGA-based laser cladding control system. This system is intended to modify the laser beam power according to the melt pool width, which is measured using a CMOS camera. All the control and monitoring tasks are carried out by a FPGA, taking advantage of its abundance of resources and speed of operation. The robustness of the image processing algorithm is assessed, as well as the control system performance. Laser power is decreased as substrate temperature increases, thus maintaining a constant clad width. This FPGA-based control system is integrated in an adaptive laser cladding system, which also includes an adaptive optical system that will control the laser focus distance on the fly. The whole system will constitute an efficient instrument for part repair with complex geometries and coating selective surfaces. This will be a significant step forward into the total industrial implementation of an automated industrial laser cladding process.

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

Alternate Fuel Cycle Technologies/Thorium Fuel Cycle Technology Programs. Quarterly report for period 1 April--30 June 1978

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

Vondra, B.L.

1978-08-01

Voloxidation and dissolution studies: rotary-kiln heat-transfer tests are under way using a small rotary kiln along with the development of a mathematical model to determine kiln-heat-flux profiles necessary to maintain a desired temperature gradient. The erosion/corrosion test for evaluating materials of construction is operational. Fuel from a BWR (Big Rock Point) yielded more fine solid residue on dissolution than in previous tests with PWR fuel. Two additional parametric voloxidation tests with H.B. Robinson fuel compared air vs pure oxygen atmospheres at 550{sup 0}C; overall tritium release and subsequent fuel dissolution were equivalent. Thorium dissolution studies: the dissolution rate of thoriamore » in fluoride-catalyzed 8 to 14 M HNO{sub 3} (100{sup 0}C) was max between 0.04 to 0.06 M HF; at higher fluoride concentrations, ThF{sub 4}.5H{sub 2}O precipitated. The rate of zircaloy dissolution continued to increase with increasing fluoride concentration. Stainless-steel-clad (Th,U)0{sub 2} fuel rods irradiated in the NRX reactor were sheared, voloxidized, and dissolved. {le}10% of the tritium was released during voloxidation in air at 600{sup 0}C. Carbon-14 removal from off-gas and fixation: carbon dioxide removal with Linde 13X molecular sieves to less than 100 ppB was experimentally verified using 300 ppM CO in air. Decontamination factors from 3000 to 7500 were obtained for CO{sub 2} removal in the gas-slurry stirred-tank reactor with CA(OH){sub 2}.or Ba(0H){sub 2}/sup .8H2O./. With Ba(OH){sub 2}.H{sub 2}0{sup 2} in a fixed-bed column, decontamination factors of about 30,000 were obtained.« less

Process of welding gamma prime-strengthened nickel-base superalloys

DOEpatents

Speigel, Lyle B.; White, Raymond Alan; Murphy, John Thomas; Nowak, Daniel Anthony

2003-11-25

A process for welding superalloys, and particularly articles formed of gamma prime-strengthened nickel-base superalloys whose chemistries and/or microstructures differ. The process entails forming the faying surface of at least one of the articles to have a cladding layer of a filler material. The filler material may have a composition that is different from both of the articles, or the same as one of the articles. The cladding layer is machined to promote mating of the faying surfaces, after which the faying surfaces are mated and the articles welded together. After cooling, the welded assembly is free of thermally-induced cracks.

Metal Matrix Composite Material by Direct Metal Deposition

NASA Astrophysics Data System (ADS)

Novichenko, D.; Marants, A.; Thivillon, L.; Bertrand, P. H.; Smurov, I.

Direct Metal Deposition (DMD) is a laser cladding process for producing a protective coating on the surface of a metallic part or manufacturing layer-by-layer parts in a single-step process. The objective of this work is to demonstrate the possibility to create carbide-reinforced metal matrix composite objects. Powders of steel 16NCD13 with different volume contents of titanium carbide are tested. On the base of statistical analysis, a laser cladding processing map is constructed. Relationships between the different content of titanium carbide in a powder mixture and the material microstructure are found. Mechanism of formation of various precipitated titanium carbides is investigated.

BWR Steam Dryer Alternating Stress Assessment Procedures

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

Morante, R. J.; Hambric, S. A.; Ziada, S.

2016-12-01

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

Thermo-mechanical assessment of full SiC/SiC composite cladding for LWR applications with sensitivity analysis

NASA Astrophysics Data System (ADS)

Singh, Gyanender; Terrani, Kurt; Katoh, Yutai

2018-02-01

SiC/SiC composites are considered among leading candidates for accident tolerant fuel cladding in light water reactors. However, when SiC-based materials are exposed to neutron irradiation, they experience significant changes in dimensions and physical properties. Under a large heat flux application (i.e. fuel cladding), the non-uniform changes in the dimensions and physical properties will lead to build-up of stresses in the structure over the course of time. To ensure reliable and safe operation of such a structure it is important to assess its thermo-mechanical performance under in-reactor conditions of irradiation and elevated temperature. In this work, the foundation for 3D thermo-mechanical analysis of SiC/SiC cladding is put in place and a set of analyses with simplified boundary conditions has been performed. The analyses were carried out with two different codes that were benchmarked against one another and prior results in the literature. A constitutive model is constructed and solved numerically to predict the stress distribution and variation in the cladding under normal operating conditions. The dependence of dimensions and physical properties variation with irradiation and temperature has been incorporated. These robust models may now be modified to take into account the axial and circumferential variation in neutron and heat flux to fully account for 3D effects. The results from the simple analyses show the development of high tensile stresses especially in the circumferential and axial directions at the inner region of the cladding. Based on the results obtained, design guidelines are recommended. For lack of certainty in or tailor-ability for the physical and mechanical properties of SiC/SiC composite material a sensitivity analysis is conducted. The analysis results establish a precedence order of the properties based on the extent to which these properties influence the temperature and the stresses.

Online monitoring of thermo-cycles and its correlation with microstructure in laser cladding of nickel based super alloy

NASA Astrophysics Data System (ADS)

Muvvala, Gopinath; Patra Karmakar, Debapriya; Nath, Ashish Kumar

2017-01-01

Laser cladding, basically a weld deposition technique, is finding applications in many areas including surface coatings, refurbishment of worn out components and generation of functionally graded components owing to its various advantages over conventional methods like TIG, PTA etc. One of the essential requirements to adopt this technique in industrial manufacturing is to fulfil the increasing demand on product quality which could be controlled through online process monitoring and correlating the signals with the mechanical and metallurgical properties. Rapid thermo-cycle i.e. the fast heating and cooling rates involved in this process affect above properties of the deposited layer to a great extent. Therefore, the current study aims to monitor the thermo-cycles online, understand its variation with process parameters and its effect on different quality aspects of the clad layer, like microstructure, elemental segregations and mechanical properties. The effect of process parameters on clad track geometry is also studied which helps in their judicious selection to deposit a predefined thickness of coating. In this study Inconel 718, a nickel based super alloy is used as a clad material and AISI 304 austenitic steel as a substrate material. The thermo-cycles during the cladding process were recorded using a single spot monochromatic pyrometer. The heating and cooling rates were estimated from the recorded thermo-cycles and its effects on microstructures were characterised using SEM and XRD analyses. Slow thermo-cycles resulted in severe elemental segregations favouring Laves phase formation and increased γ matrix size which is found to be detrimental to the mechanical properties. Slow cooling also resulted in termination of epitaxial growth, forming equiaxed grains near the surface, which is not preferred for single crystal growth. Heat treatment is carried out and the effect of slow cooling and the increased γ matrix size on dissolution of segregated elements in metal matrix is studied.

Steam Oxidation Testing in the Severe Accident Test Station

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

Pint, Bruce A.; McMurray, Jake W.

2016-08-01

Since 2011, Oak Ridge National Laboratory (ORNL) has been conducting high temperature steam oxidation testing of candidate alloys for accident tolerant fuel (ATF) cladding. These concepts are designed to enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the US ATF community, the Severe Accident Test Station (SATS) has been evaluating candidate materials (including coatings) since 2012. Compared to the current UO 2/Zr-based alloy fuel system, alternative cladding materials need to offer slower oxidation kinetics and a smaller enthalpy of oxidation in order to significantly reduce the rate of heat and hydrogen generation in the coremore » during a coolant-limited severe accident. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models. However, prior modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. In some cases, the results have been difficult to interpret and more fundamental information is needed such as the stability of alumina in flowing steam at 1400°-1500°C. This report summarizes recent work to measure the steam oxidation kinetics of candidate alloys, the evaporation rate of alumina in steam and the development of integral data on FeCrAl compared to conventional Zr-based cladding.« less

Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux

NASA Astrophysics Data System (ADS)

Petrie, Christian M.; Koyanagi, Takaaki; McDuffee, Joel L.; Deck, Christian P.; Katoh, Yutai; Terrani, Kurt A.

2017-08-01

The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300-350 °C under a representative high heat flux (∼0.66 MW/m2) during one cycle of irradiation in an un-instrumented ;rabbit; capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb the expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. The success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.

Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux

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

Petrie, Christian M.; Koyanagi, Takaaki; McDuffee, Joel L.

The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300–350 °C under a representative high heat flux (~0.66 MW/m 2) during one cycle of irradiation in an un-instrumented “rabbit” capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb themore » expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. Furthermore, the success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.« less

Use of the Hugoniot elastic limit in laser shockwave experiments to relate velocity measurements

NASA Astrophysics Data System (ADS)

Smith, James A.; Lacy, Jeffrey M.; Lévesque, Daniel; Monchalin, Jean-Pierre; Lord, Martin

2016-02-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. This fuel-cladding interface qualification will ensure the survivability of the fuel plates in the harsh reactor environment even under abnormal operating conditions. One of the concerns of the project is the difficulty of calibrating and standardizing the laser shock technique. An analytical study under development and experimental testing supports the hypothesis that the Hugoniot Elastic Limit (HEL) in materials can be a robust and simple benchmark to compare stresses generated by different laser shock systems.

Stress corrosion crack initiation of Zircaloy-4 cladding tubes in an iodine vapor environment during creep, relaxation, and constant strain rate tests

NASA Astrophysics Data System (ADS)

Jezequel, T.; Auzoux, Q.; Le Boulch, D.; Bono, M.; Andrieu, E.; Blanc, C.; Chabretou, V.; Mozzani, N.; Rautenberg, M.

2018-02-01

During accidental power transient conditions with Pellet Cladding Interaction (PCI), the synergistic effect of the stress and strain imposed on the cladding by thermal expansion of the fuel, and corrosion by iodine released as a fission product, may lead to cladding failure by Stress Corrosion Cracking (SCC). In this study, internal pressure tests were conducted on unirradiated cold-worked stress-relieved Zircaloy-4 cladding tubes in an iodine vapor environment. The goal was to investigate the influence of loading type (constant pressure tests, constant circumferential strain rate tests, or constant circumferential strain tests) and test temperature (320, 350, or 380 °C) on iodine-induced stress corrosion cracking (I-SCC). The experimental results obtained with different loading types were consistent with each other. The apparent threshold hoop stress for I-SCC was found to be independent of the test temperature. SEM micrographs of the tested samples showed many pits distributed over the inner surface, which tended to coalesce into large pits in which a microcrack could initiate. A model for the time-to-failure of a cladding tube was developed using finite element simulations of the viscoplastic mechanical behavior of the material and a modified Kachanov's damage growth model. The times-to-failure predicted by this model are consistent with the experimental data.

High Sensitive Temperature Sensor Using a Liquid-core Optical Fiber with Small Refractive Index Difference Between Core and Cladding Materials.

PubMed

Xu, Yonghao; Chen, Xianfeng; Zhu, Yu

2008-03-17

An intensive temperature sensor based on a liquid-core optical fiber has been demonstrated for the measuring the temperature of the environment. The core of fiber is filled with a mixture of toluene and chloroform in order to make the refractive index of the liquid-core and the cladding of the fiber close. The experiment shows that a temperature sensitivity of about 5 dB/K and a tunable temperature range (from 20 o C to 60 o C) can be achieved. Based on the dielectric-clad liquid core fiber model, a simulation was carried out and the calculated results were in good accord with the experimental measurement.

Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

NASA Astrophysics Data System (ADS)

Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

2018-02-01

The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

Single-polarization hollow-core square photonic bandgap waveguide

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

Eguchi, Masashi, E-mail: megu@ieee.org; Tsuji, Yasuhide, E-mail: y-tsuji@mmm.muroran-it.ac.jp

Materials with a periodic structure have photonic bandgaps (PBGs), in which light can not be guided within certain wavelength ranges; thus light can be confined within a low-index region by the bandgap effect. In this paper, rectangular-shaped hollow waveguides having waveguide-walls (claddings) using the PBG have been discussed. The design principle for HE modes of hollow-core rectangular PBG waveguides with a Bragg cladding consisting of alternating high- and low-index layers, based on a 1D periodic multilayer approximation for the Bragg cladding, is established and then a novel single-polarization hollow-core square PBG waveguide using the bandgap difference between two polarized wavesmore » is proposed. Our results demonstrated that a single-polarization guiding can be achieved by using the square Bragg cladding structure with different layer thickness ratios in the mutually orthogonal directions and the transmission loss of the guided mode in a designed hollow-core square PBG waveguide is numerically estimated to be 0.04 dB/cm.« less

Compatibility studies of metallic materials with lithium-based oxides

NASA Astrophysics Data System (ADS)

Hofmann, P.; Dienst, W.

1988-07-01

The compatibility of Li 2O, Li 4SiO 4 and Li 2SiO 3 with the cladding materials AISI 316, 1.4914, Hastelloy X and Inconel 625 was investigated at 800-1000°C for annealing times up to 1000 h. A controlled oxygen reactivity was established by adding 1 mol% NiO per mole Li 2O to the Li-based oxides. In addition, some compatibility tests were performed at 600-900°C on Be, which is of interest as a neutron multiplier material, with Li 2SiO 3 as well as AISI 316. Li 2O accounted for the strongest cladding attack, followed by Li 4SiO 4 and Li 2SiO 3. In the absence of NiO, Li 2SiO 3 caused no chemical interactions at all. With respect to the cladding materials, there was no considerable difference in the reaction rates of AISI 316, Hastelloy X and Inconel 625. However, the steel 1.4914 was clearly more heavily attacked at and above 800°C. The compatibility of Be with Li 2SiO 3 or AISI 316 seems to be tolerable up to about 650°C. At higher temperatures a liquid Li suicide phase is formed which results in strong local attack and penetration into Li 2SiO 3.

PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

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

Hales, J. D.; Gamble, K. A.

2015-09-01

Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normalmore » operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.« less

Investigation of semiconductor clad optical waveguides

NASA Technical Reports Server (NTRS)

Batchman, T. E.; Carson, R. F.

1985-01-01

A variety of techniques have been proposed for fabricating integrated optical devices using semiconductors, lithium niobate, and glasses as waveguides and substrates. The use of glass waveguides and their interaction with thin semiconductor cladding layers was studied. Though the interactions of these multilayer waveguide structures have been analyzed here using glass, they may be applicable to other types of materials as well. The primary reason for using glass is that it provides a simple, inexpensive way to construct waveguides and devices.

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

Walston, S; Rowland, M; Campbell, K

It is difficult to track to the location of a melted core in a GE BWR with Mark I containment during a beyond-design-basis accident. The Cooper Nuclear Station provided a baseline of normal material distributions and shielding configurations for the GE BWR with Mark I containment. Starting with source terms for a design-basis accident, methods and remote observation points were investigated to allow tracking of a melted core during a beyond-design-basis accident. The design of the GE BWR with Mark-I containment highlights an amazing poverty of expectations regarding a common mode failure of all reactor core cooling systems resulting inmore » a beyond-design-basis accident from the simple loss of electric power. This design is shown in Figure 1. The station blackout accident scenario has been consistently identified as the leading contributor to calculated probabilities for core damage. While NRC-approved models and calculations provide guidance for indirect methods to assess core damage during a beyond-design-basis loss-of-coolant accident (LOCA), there appears to be no established method to track the location of the core directly should the LOCA include a degree of fuel melt. We came to the conclusion that - starting with detailed calculations which estimate the release and movement of gaseous and soluble fission products from the fuel - selected dose readings in specific rooms of the reactor building should allow the location of the core to be verified.« less

Modified rod-in-tube for high-NA tellurite glass fiber fabrication: materials and technologies.

PubMed

Chen, Qiuling; Wang, Hui; Wang, Qingwei; Chen, Qiuping; Hao, Yinlei

2015-02-01

In this paper, we report the whole fabrication process for high-numerical aperture (NA) tellurite glass fibers from material preparation to preform fabrication, and eventually, fiber drawing. A tellurite-based high-NA (0.9) magneto-optical glass fiber was drawn successfully and characterized. First, matchable core and cladding glasses were fabricated and matched in terms of physical properties. Second, a uniform bubble-free preform was fabricated by means of a modified rod-in-tube technique. Finally, the fiber drawing process was studied and optimized. The high-NA fibers (∅(core), 40-50 μm and ∅(cladding), 120-130 μm) so obtained were characterized for their geometrical and optical properties.

Silicon carbide composite for light water reactor fuel assembly applications

NASA Astrophysics Data System (ADS)

Yueh, Ken; Terrani, Kurt A.

2014-05-01

The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

Boiling-Water Reactor internals aging degradation study. Phase 1

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

Luk, K.H.

1993-09-01

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor drymore » tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.« less

Temperature dependent absorption measurement of various transition metal doped laser materials

NASA Astrophysics Data System (ADS)

Horackova, Lucie; Šulc, Jan; Jelinkova, Helena; Jambunathan, Venkatesan; Lucianetti, Antonio; Mocek, Tomás.

2015-05-01

In recent years, there has been a vast development of high energy class lasers of the order of 100 J to kJ level which have potential applications in the field of science and technology. Many such systems use the gain media cooled at cryogenic temperatures which will help in enhancing the spectroscopic and thermo-optical properties. Nevertheless, parasitic effects like amplified spontaneous emission enhance and affect the overall efficiency. The best way to suppress this effect is to use cladding element attached to the gain material. Based on these facts, this work was focused on the systematic investigation of temperature dependent absorption of several materials doped with transition metals, which can be used as cladding, as laser gain material, or as passive Q-switching element. The Ti:sapphire, Cr:YAG, V:YAG, and Co:MALO samples were measured in temperature range from 80 K to 330 K by step of 50 K. Using Beer-Lambert law we estimated the absorption coefficient of these materials.

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

Initial experimental evaluation of crud-resistant materials for light water reactors

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Carlson, M. B.; Hussey, D.; Short, M. P.

2018-01-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud causes serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each assumes that it will always be present. In this study, we report on the development of crud-resistant materials as fuel cladding coatings, to reduce or eliminate these problems altogether. Integrated loop testing experiments at flowing LWR conditions show significantly reduced crud adhesion and surface crud coverage, respectively, for TiC and ZrN coatings compared to ZrO2. The loop testing results roughly agree with the London dispersion component of van der Waals force predictions, suggesting that they contribute most significantly to the adhesion of crud to fuel cladding in out-of-pile conditions. These results motivate a new look at ways of reducing crud, thus avoiding many expensive LWR operational issues.

Purification of nuclear grade Zr scrap as the high purity dense Zr deposits from Zirlo scrap by electrorefining in LiF-KF-ZrF4 molten fluorides

NASA Astrophysics Data System (ADS)

Park, Kyoung Tae; Lee, Tae Hyuk; Jo, Nam Chan; Nersisyan, Hayk H.; Chun, Byong Sun; Lee, Hyuk Hee; Lee, Jong Hyeon

2013-05-01

Zirconium (Zr) has commonly been used as a cladding material of nuclear fuel. Moreover, it is regarded as the only material that can be used for nuclear fuel cladding because it has the lowest neutron capture cross section of any metal element and because it has high corrosion resistance and size stability. In this study, Hf-free Zr tubes (Zr-1Nb-1Sn-0.1Fe) were used as anode materials and electrorefining was performed in a LiF-KF eutectic 6 wt.% ZrF4 molten fluoride salt system. As a result of electrolysis, Zr scrap metal was recycled into pure Zr with low levels of impurities, and the size and density of the Zr deposit was controlled using applied current density.

Equations of state for crystalline zirconium iodide: The role of dispersion

NASA Astrophysics Data System (ADS)

Rossi, Matthew L.; Taylor, Christopher D.

2013-02-01

We present the first-principle equations of state of several zirconium iodides, ZrI2, ZrI3, and ZrI4, computed using density functional theory methods that apply various methods for introducing the dispersion correction. Iodides formed due to reaction of molecular or atomic iodine with zirconium and zircaloys are of particular interest due to their application to the cladding material used in the fabrication of nuclear fuel rods. Stress corrosion cracking (SCC), associated with fission product chemistry with the clad material, is a major concern in the life cycle of nuclear fuels, as many of the observed rod failures have occurred due to pellet-cladding chemical interactions (PCCI) [A. Atrens, G. Dannhäuser, G. Bäro, Stress-corrosion-cracking of zircaloy-4 cladding tubes, Journal of Nuclear Materials 126 (1984) 91-102; P. Rudling, R. Adamson, B. Cox, F. Garzarolli, A. Strasser, High burn-up fuel issues, Nuclear Engineering and Technology 40 (2008) 1-8]. A proper understanding of the physical properties of the corrosion products is, therefore, required for the development of a comprehensive SCC model. In this particular work, we emphasize that, while existing modeling techniques include methods to compute crystal structures and associated properties, it is important to capture intermolecular forces not traditionally included, such as van der Waals (dispersion) correction. Furthermore, crystal structures with stoichiometries favoring a high I:Zr ratio are found to be particularly sensitive, such that traditional density functional theory approaches that do not incorporate dispersion incorrectly predict significantly larger volumes of the lattice. This latter point is related to the diffuse nature of the iodide electron cloud.

Theoretical and experimental study of bent fully aperiodic large-pitch fibers for enhancing the high-order modes delocalization

NASA Astrophysics Data System (ADS)

du Jeu, Rémi; Dauliat, Romain; Darwich, Dia; Auguste, Jean-Louis; Benoît, Aurélien; Leconte, Baptiste; Malleville, Marie-Alicia; Jamier, Raphaël.; Schuster, Kay; Roy, Philippe

2018-02-01

The power scaling of fiber lasers and amplifiers has triggered an extensive development of large-mode area fibers among which the most promising are the distributed mode filtering fibers and the large-pitch fibers. These structures enable for an effective higher-order modes delocalization and subsequently a singlemode emission. An interesting alternative consists in using the fully-aperiodic large-pitch fibers, into which the standard air-silica photonic crystal cladding is replaced by an aperiodic pattern made of solid low-index inclusions cladding. However, in such a structure, the core and the background cladding material surrounding it must have rigorously the same refractive index. Current synthesis processes and measurement techniques offer respectively a maximum resolution of 5×10-4 and 1×10-4 while the indexmatching must be as precise as 1×10-5 . Lately a gain material with a refractive index 1.5×10-4 higher than that of the background cladding material was fabricated, thus re-confining the first higher-order modes in the core. A numerical study is carried out on the benefit of bending such fully-aperiodic fiber to counteract this phenomenon. Optimized bending axis and radius have been determined. Experiments are done in a laser cavity operating at 1030 nm using an 88cm-long 51μm core diameter ytterbium-doped fiber. Results demonstrate an improvement of the M2 from 1.7 when the fiber is kept straight to 1.2 when it is bent with a 100 to 60 cm bend radius. These primary results are promising for future power scaling.

The effect of zinc injection on the increasing of Inconel 600 TT corrosion resistances

NASA Astrophysics Data System (ADS)

Febrianto; Sriyono; Widodo, Surip; Sunaryo, Geni Rina

2018-02-01

Many failures were found in reactor pressure vessel head penetration (RPV) head material. Those failures caused by boric acid corrosion, and from visual examination were found a big hole and white deposit crystal of boric acid during shutdown maintenance at David Besse reactor. Zinc Oxide addition in BWR reactor known as Zinc Injection that has purposed to reduce radiation exposure cause of Hydrogen addition. Beside reducing the radiation exposure, Zinc injection also has an effect in reducing material corrosion. The purpose of study is to determine the effect of zinc addition, boric acid, temperature also the effects of Cobalt Nitrate and Zinc Oxide addition to Inconel 600 TT as RPV head penetration material. The result in the BWR reactor experience will be implementated at PWR reactor, weather zinc oxide addition also has an effect in reducing the corrosion of Inconel 600. The method that used in this research is to observe the corrosion rates for Inconel 600 material using Potentiostat. Examination were conducted in 30, 40, 60, 70, 80 and 80 °C using 1000, 1500, 2000, 2500 and 3000 ppm boric acid concentration. The results showed that the corrosion rate for the material were very small, but the highest corrosion rate occurred in 3000 ppm boric acid concentration at 90 °C with Cobalt Nitrate addition, around 5.210 x 10-1 mpy. In the same condition at 3000 ppm boric acid concentration for temperature at 90 °C, Inconel 600 TT corrosion rate is smaller with Zinc oxide addition, around 4.631 x 10-1 mpy.

Proposal to Produce Novel, Transparent Radiation Hard Low Refractive Index

DTIC Science & Technology

1994-02-09

or ainy nht.at NX ftU 1.AECY USE a EP....3RFPfT TYP’E AND DATES~ COV-ERED-- 4. TITLE AND SUBTITLE . . FjUNDING NUMBERS PROPOSAL TO PRODUCE NOVEL...cladding use . our research resulted in identifying a radiation hard, low refractive index polymer, poly (heptafluorobutyl methacrylate), P(MFBM) as the best...candidate for a novel ~. cladding material. P(HFB) has a refractive index of 1.387. When used to clada styrene core, the theoretical light propagation

Fineblanking, Diffusion Bonding, and Testing of Fluidic Laminates.

DTIC Science & Technology

1980-07-01

AD-AU69 347 TRITEC INC COLUMBIA ND F/$ 13/7 FINEBLANKING, DIFFUSION BONDING, AND TESTING OF FLUIDIC LAMINAT --ETCIU) JUL 80 L K PECAN OAAK21-79-C-0074...amplifier assembly. The effects of die roll and burrs can be minimized by secondary operations *such as abrasive machining , but this adds to the expense...clad material. Experience has shown that a clad thickness of 0.038 + 0.008 mm is required for the semi-solid diffusion bonding process. The composition

Reversed Cherenkov emission of terahertz waves from an ultrashort laser pulse in a sandwich structure with nonlinear core and left-handed cladding.

PubMed

Bakunov, M I; Mikhaylovskiy, R V; Bodrov, S B; Luk'yanchuk, B S

2010-01-18

We propose a scheme for an experimental verification of the reversed Cherenkov effect in left-handed media. The scheme uses optical-to-terahertz conversion in a planar sandwichlike structure that consists of a nonlinear core cladded with a material that exhibits left-handedness at terahertz frequencies. The focused into a line femtosecond laser pulse propagates in the core and emits Cherenkov wedge of terahertz waves in the cladding. We developed a theory that describes terahertz generation in such a structure and calculated spatial distribution of the generated terahertz field, its energy spectrum, and optical-to-terahertz conversion efficiency. The proposed structure can be a useful tool for characterization of the electromagnetic properties of metamaterials in the terahertz frequency range.

Sediment Budgeting in Dam-Affected Rivers: Assessing the Influence of Damming, Tributaries, and Alluvial Valley Sediment Storage on Sediment Regimes

NASA Astrophysics Data System (ADS)

Wilcox, A. C.; Dekker, F. J.; Riebe, C. S.

2014-12-01

Although sediment supply is recognized as a fundamental driver of fluvial processes, measuring how dams affect sediment regimes and incorporating such knowledge into management strategies remains challenging. To determine the influences of damming, tributary supply, and valley morphology and sediment storage on downstream sediment supply in a dryland river, the Bill Williams River (BWR) in western Arizona, we measured basin erosion rates using cosmogenic nuclide analysis of beryllium-10 (10Be) at sites upstream and downstream of a dam along the BWR, as well as from tributaries downstream of the dam. Riverbed sediment mixing calculations were used to test if the dam, which blocks sediment supply from the upper 85% of the basin's drainage area, increases the proportion of tributary sediment to residual upstream sediment in mainstem samples downstream of the dam. Erosion rates in the BWR watershed are more than twice as large in the upper catchment (136 t km-2 yr-1) than in tributaries downstream of Alamo Dam (61 t km-2 yr-1). Tributaries downstream of the dam have little influence on mainstem sediment dynamics. The effect of the dam on reducing sediment supply is limited, however, because of the presence of large alluvial valleys along the mainstem BWR downstream of the dam that store substantial sediment and mitigate supply reductions from the upper watershed. These inferences, from our 10Be derived erosion rates and mixing calculations, are consistent with field observations of downstream changes in bed material size, which suggest that sediment-deficit conditions are restricted to a 10 km reach downstream of the dam, and limited reservoir bathymetry data. Many studies have suggested that tributary sediment inputs downstream of dams play a key role in mitigating dam-induced sediment deficits, but here we show that in a dryland river with ephemeral tributaries, sediment stored in alluvial valleys can also play a key role and in some cases trumps the role of tributaries.

Mechanistic Considerations Used in the Development of the PROFIT PCI Failure Model

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

Pankaskie, P. J.

A fuel Pellet-Zircaloy Cladding (thermo-mechanical-chemical) Interactions (PC!) failure model for estimating the probability of failure in !ransient increases in power (PROFIT) was developed. PROFIT is based on 1) standard statistical methods applied to available PC! fuel failure data and 2) a mechanistic analysis of the environmental and strain-rate-dependent stress versus strain characteristics of Zircaloy cladding. The statistical analysis of fuel failures attributable to PCI suggested that parameters in addition to power, transient increase in power, and burnup are needed to define PCI fuel failures in terms of probability estimates with known confidence limits. The PROFIT model, therefore, introduces an environmentalmore » and strain-rate dependent strain energy absorption to failure (SEAF) concept to account for the stress versus strain anomalies attributable to interstitial-disloction interaction effects in the Zircaloy cladding. Assuming that the power ramping rate is the operating corollary of strain-rate in the Zircaloy cladding, then the variables of first order importance in the PCI fuel failure phenomenon are postulated to be: 1. pre-transient fuel rod power, P{sub I}, 2. transient increase in fuel rod power, {Delta}P, 3. fuel burnup, Bu, and 4. the constitutive material property of the Zircaloy cladding, SEAF.« less

Fabrication and evaluation of evanescent wave absorption based polyaniline-cladding modified fiber optic urea biosensor

NASA Astrophysics Data System (ADS)

Botewad, S. N.; Pahurkar, V. G.; Muley, G. G.

2018-01-01

The fabrication and study of cladding modified intrinsic fiber optic urea biosensor has been reported in the present investigation. A simple cladding modification technique was used to construct the sensor by uncladding the small portion from optical fiber. Further bare core was decorated by supportive porous, chemically and optically sensitive matrix material polyaniline (PANI) as an active cladding for enzyme residency. Enzyme-urease (Urs) was cross-linked on the active cladding region via glutaraldehyde solution. Confirmation of the prepared PANI in proper form determined by ultraviolet-visible and Fourier transform infrared spectroscopic techniques. X-ray diffraction technique was employed for nature and compatibility examination of PANI. Sensor parameters such as sensitivity, selectivity, stability and lower detection limit have been analyzed by absorption variation study in evanescent wave field. The response of prepared sensor was studied towards urea in the wide concentration range 100 nM-100 mM and confirmed its lowest detection limit as 100 nM. The stability of sensor was found 28 days with little variation in response. The fabricated sensor has not shown any response towards interference species like glucose, ascorbic acid, L-alanine, L-arginine and their combination with urea solution and hence found selective for urea solution only.

WRC bulletin. A review of underclad cracking in pressure-vessel components

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

Vinckier, A.G.; Pense, A.W.

1974-01-01

This review of cracking underneath the weld cladding is to determine what factors contribute to this condition, and to outline means for alleviating or eliminating this condition. Considerable data on manufacture, heat treatment, and cladding of heavy-section pressure-vessel steels for nuclear service are also included. Three factors in combination that promote underclad cracking are susceptible microstructure, favorable residual-stress pattern, and a thermal treatment bringing the steel into a critical temperature region (600-650/sup 0/C) where creep ductility is low. High-heat-input weld-overlay cladding produces the susceptible microstructure and residual-stress pattern and postweld heat treatment produces the critical temperature. Most underclad cracking wasmore » found in SA508 Class 2 steel forgings clad with one-layer submerged-arc strip electrodes or multi-electrode processes. It was not produced in SA533 Grade B plate or when multilayer overlay processes were used. Underclad cracking can be reduced or eliminated by a two-layer cladding technique, by controlling welding process variables (low heat input), renormalizing the sensitive HAZ region prior to heat treatment, by use of nonsusceptible materials, or by eliminating the postweld heat treatment. Results of a questionnaire survey are also included. 50 references. (DLC)« less

JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

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

G. Borges

2006-01-31

The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long termmore » microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.« less

Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

2011-06-01

Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent plastic strains are reduced; and (3) the maximum first principal stresses for certain burnup at the matrix or the cladding are lower than the ones without the hardening effect, and the differences are found to increase with burnup; and the variation rules of the interfacial stresses are similar.

Hydride reorientation and its impact on ambient temperature mechanical properties of high burn-up irradiated and unirradiated recrystallized Zircaloy-2 nuclear fuel cladding with an inner liner

NASA Astrophysics Data System (ADS)

Auzoux, Q.; Bouffioux, P.; Machiels, A.; Yagnik, S.; Bourdiliau, B.; Mallet, C.; Mozzani, N.; Colas, K.

2017-10-01

Precipitation of radial hydrides in zirconium-based alloy cladding concomitant with the cooling of spent nuclear fuel during dry storage can potentially compromise cladding integrity during its subsequent handling and transportation. This paper investigates hydride reorientation and its impact on ductility in unirradiated and irradiated recrystallized Zircaloy-2 cladding with an inner liner (cladding for boiling water reactors) subjected to hydride reorientation treatments. Cooling from 400 °C, hydride reorientation occurs in recrystallized Zircaloy-2 with liner at a lower effective stress in irradiated samples (below 40 MPa) than in unirradiated specimens (between 40 and 80 MPa). Despite significant hydride reorientation, unirradiated recrystallized Zircaloy-2 with liner cladding containing ∼200 wppm hydrogen shows a high diametral strain at fracture (>15%) during burst tests at ambient temperature. This ductile behavior is due to (1) the lower yield stress of the recrystallized cladding materials in comparison to hydride fracture strength (corrected by the compression stress arising from the precipitation) and (2) the hydride or hydrogen-depleted zone as a result of segregation of hydrogen into the liner layer. In irradiated Zircaloy-2 with liner cladding containing ∼340 wppm hydrogen, the conservation of some ductility during ring tensile tests at ambient temperature after reorientation treatment at 400 °C with cooling rates of ∼60 °C/h is also attributed to the existence of a hydride-depleted zone. Treatments at lower cooling rates (∼6 °C/h and 0.6 °C/h) promote greater levels of hydrogen segregation into the liner and allow for increased irradiation defect annealing, both of which result in a significant increase in ductility. Based on this investigation, given the very low cooling rates typical of dry storage systems, it can be concluded that the thermal transients associated with dry storage should not degrade, and more likely should actually improve, ductility of recrystallized Zircaloy-2 cladding with inner liner with such hydrogen content.

Structure and properties of steel case-hardened by non-vacuum electron-beam cladding of carbon fibers

NASA Astrophysics Data System (ADS)

Losinskaya, A. A.; Lozhkina, E. A.; Bardin, A. I.

2017-12-01

At the present time, the actual problem of materials science is the increase in the steels performance characteristics. In the paper some mechanical properties of the case-hardened materials received by non-vacuum electron-beam cladding of carbon fibers are determined. The depth of the hardened layers varies from 1.5 to 3 mm. The impact strength of the samples exceeds 50 J/cm2. The wear resistance of the coatings obtained exceeds the properties of steel 20 after cementation and quenching with low tempering. The results of a study of the microhardness of the resulting layers and the microstructure are also given. The hardness of the surface layers exceeds 5700 MPa.

Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

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

Clayton, Dwight A.; Poore, III, Willis P.

2015-01-01

The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Markmore » I plant for those instrumentation systems considered most important for accident management purposes.« less

Modification of tribology and high-temperature behavior of Ti 48Al 2Cr 2Nb intermetallic alloy by laser cladding

NASA Astrophysics Data System (ADS)

Liu, Xiu-Bo; Wang, Hua-Ming

2006-06-01

In order to improve the tribology and high-temperature oxidation properties of the Ti-48Al-2Cr-2Nb intermetallic alloy simultaneously, mixed NiCr-Cr 3C 2 precursor powders had been investigated for laser cladding treatment to modify wear and high-temperature oxidation resistance of the material. The alloy samples were pre-placed with NiCr-80, 50 and 20%Cr 3C 2 (wt.%), respectively, and laser treated at the same parameters, i.e., laser output power 2.8 kW, beam scanning speed 2.0 mm/s, beam dimension 1 mm × 18 mm. The treated samples underwent tests of microhardness, wear and high-temperature oxidation. The results showed that laser cladding with different constitution of mixed precursor NiCr-Cr 3C 2 powders improved surface hardness in all cases. Laser cladding with NiCr-50%Cr 3C 2 resulted in the best modification of tribology and high-temperature oxidation behavior. X-ray diffraction (XRD), optical microscope (OM), scanning electron microscopy (SEM) and energy-dispersive spectrometer (EDS) analyses indicated that the formation of reinforced Cr 7C 3, TiC and both continuous and dense Al 2O 3, Cr 2O 3 oxide scales were supposed to be responsible for the modification of the relevant properties. As a result, the present work had laid beneficial surface engineering foundation for TiAl alloy applied as future light weight and high-temperature structural candidate materials.

Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement.

PubMed

Škarohlíd, Jan; Ashcheulov, Petr; Škoda, Radek; Taylor, Andrew; Čtvrtlík, Radim; Tomáštík, Jan; Fendrych, František; Kopeček, Jaromír; Cháb, Vladimír; Cichoň, Stanislav; Sajdl, Petr; Macák, Jan; Xu, Peng; Partezana, Jonna M; Lorinčík, Jan; Prehradná, Jana; Steinbrück, Martin; Kratochvílová, Irena

2017-07-25

In this work, we demonstrate and describe an effective method of protecting zirconium fuel cladding against oxygen and hydrogen uptake at both accident and working temperatures in water-cooled nuclear reactor environments. Zr alloy samples were coated with nanocrystalline diamond (NCD) layers of different thicknesses, grown in a microwave plasma chemical vapor deposition apparatus. In addition to showing that such an NCD layer prevents the Zr alloy from directly interacting with water, we show that carbon released from the NCD film enters the underlying Zr material and changes its properties, such that uptake of oxygen and hydrogen is significantly decreased. After 100-170 days of exposure to hot water at 360 °C, the oxidation of the NCD-coated Zr plates was typically decreased by 40%. Protective NCD layers may prolong the lifetime of nuclear cladding and consequently enhance nuclear fuel burnup. NCD may also serve as a passive element for nuclear safety. NCD-coated ZIRLO claddings have been selected as a candidate for Accident Tolerant Fuel in commercially operated reactors in 2020.

Investigation of the structure and properties of boron-containing coatings obtained by electron-beam treatment

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

Krivezhenko, Dina S., E-mail: dinylkaa@yandex.ru; Drobyaz, Ekaterina A., E-mail: ekaterina.drobyaz@yandex.ru; Bataev, Ivan A., E-mail: ivanbataev@ngs.ru

2015-10-27

An investigation of surface-hardened materials obtained by cladding with an electron beam injected into the air atmosphere was carried out. Structural investigations of coatings revealed that an increase in boron carbide concentration in a saturating mixture contributed to a rise of a volume fraction of iron borides in coatings. The maximum hardened depth reached 2 mm. Hardened layers were characterized by the formation of heterogeneous structure which consisted of iron borides and titanium carbides distributed uniformly in the eutectic matrix. Areas of titanium boride conglomerations were detected. It was found that an increase in the boron carbide content led to anmore » enhancement in hardness of the investigated materials. Friction testing against loosely fixed abrasive particles showed that electron-beam cladding of powder mixtures containing boron carbides, titanium, and iron in air atmosphere allowed enhancing a resistance of materials hardened in two times.« less

Reduced yield stress for zirconium exposed to iodine: Reactive force field simulation

DOE PAGES

Rossi, Matthew L.; Taylor, Christopher D.; van Duin, Adri C. T.

2014-11-04

Iodine-induced stress-corrosion cracking (ISCC), a known failure mode for nuclear fuel cladding, occurs when iodine generated during the irradiation of a nuclear fuel pellet escapes the pellet through diffusion or thermal cracking and chemically interacts with the inner surface of the clad material, inducing a subsequent effect on the cladding’s resistance to mechanical stress. To complement experimental investigations of ISCC, a reactive force field (ReaxFF) compatible with the Zr-I chemical and materials systems has been developed and applied to simulate the impact of iodine exposure on the mechanical strength of the material. The study shows that the material’s resistance tomore » stress (as captured by the yield stress of a high-energy grain boundary) is related to the surface coverage of iodine, with the implication that ISCC is the result of adsorption-enhanced decohesion.« less

Investigation of the structure and properties of boron-containing coatings obtained by electron-beam treatment

NASA Astrophysics Data System (ADS)

Krivezhenko, Dina S.; Drobyaz, Ekaterina A.; Bataev, Ivan A.; Chuchkova, Lyubov V.

2015-10-01

An investigation of surface-hardened materials obtained by cladding with an electron beam injected into the air atmosphere was carried out. Structural investigations of coatings revealed that an increase in boron carbide concentration in a saturating mixture contributed to a rise of a volume fraction of iron borides in coatings. The maximum hardened depth reached 2 mm. Hardened layers were characterized by the formation of heterogeneous structure which consisted of iron borides and titanium carbides distributed uniformly in the eutectic matrix. Areas of titanium boride conglomerations were detected. It was found that an increase in the boron carbide content led to an enhancement in hardness of the investigated materials. Friction testing against loosely fixed abrasive particles showed that electron-beam cladding of powder mixtures containing boron carbides, titanium, and iron in air atmosphere allowed enhancing a resistance of materials hardened in two times.

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

Hybrid nuclear reactor grey rod to obtain required reactivity worth

DOEpatents

Miller, John V.; Carlson, William R.; Yarbrough, Michael B.

1991-01-01

Hybrid nuclear reactor grey rods are described, wherein geometric combinations of relatively weak neutron absorber materials such as stainless steel, zirconium or INCONEL, and relatively strong neutron absorber materials, such as hafnium, silver-indium cadmium and boron carbide, are used to obtain the reactivity worths required to reach zero boron change load follow. One embodiment includes a grey rod which has combinations of weak and strong neutron absorber pellets in a stainless steel cladding. The respective pellets can be of differing heights. A second embodiment includes a grey rod with a relatively thick stainless steel cladding receiving relatively strong neutron absorber pellets only. A third embodiment includes annular relatively weak netron absorber pellets with a smaller diameter pellet of relatively strong absorber material contained within the aperture of each relatively weak absorber pellet. The fourth embodiment includes pellets made of a homogeneous alloy of hafnium and a relatively weak absorber material, with the percentage of hafnium chosen to obtain the desired reactivity worth.

Photonic-crystal fiber as a multifunctional optical sensor and sample collector.

PubMed

Konorov, Stanislav; Zheltikov, Aleksei; Scalora, Michael

2005-05-02

Two protocols of optical sensing realized with the same photonic-crystal fiber are compared. In the first protocol, diode-laser radiation is delivered to a sample through the central core of a dual-cladding photonic-crystal fiber with a diameter of a few micrometers, while the large-diameter fiber cladding serves to collect the fluorescent response from the sample and to guide it to a detector in the backward direction. In the second scheme, liquid sample is collected by a microcapillary array in the fiber cladding and is interrogated by laser radiation guided in the fiber modes. For sample fluids with refractive indices exceeding the refractive index of the fiber material, fluid channels in photonic-crystal fibers can guide laser light by total internal reflection, providing an 80% overlap of interrogating radiation with sample fluid.

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

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

Marshall, William BJ J

2016-01-01

A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blademore » histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.« less

Water chemistry control and decontamination experience with TEPCO BWR`s and the measures planned for the future

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

Suzuki, N.; Miyamaru, K.

1995-03-01

The new TEPCO BWR`s are capable of having the occupational radiation exposure controlled successfully at a low level by selecting low cobalt steel, using corrosion-resistant steel, employing dual condensate polishing systems, and controlling Ni/Fe ratio during operation. The occupational radiation exposure of the old BWR`s, on the other hand, remains high though reduced substantially through the use of low cobalt replacement steel and the partial addition of a filter in the condensate polishing system. Currently under review is the overall decontamination procedure for the old BWR`s to find out to measures needed to reduce the amount of crud that ismore » and has been carried over into the nuclear reactor. The current status of decontamination is reported below.« less

Air Vehicle Integration and Technology Research (AVIATR). Delivery Order 0002: Condition-Based Maintenance Plus Structural Integrity (CBM+SI) Strategy Development

DTIC Science & Technology

2010-11-01

material. The rubber is laser -etched with rows of tiny, interconnected channels or galleries, to which air pressure is applied. Any propagating crack... clad one side. The Upper Lobe has a radius of approximately 85” (compound curvature) in the region of interest. As stated previously, the skin is...7079-T6 sheet; clad one side with a varying thickness of 0.050” to 0.071” (varies according to stability requirements for compression combined with

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

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

Sridharan, Kumar; Mariani, Robert; Bai, Xianming

Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO 2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins inmore » accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO 4) and silica (SiO 2) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi 2 coating) during clad quenching experiments is discussed in detail.« less

Histopathological Findings Associated With Gastroesophageal Reflux Disease and Aspiration After Lung Transplantation: Initial Brazilian Single-Center Experience.

PubMed

Carraro, R M; Nascimento, E C T; Szachnowicz, S; Camargo, P C L B; Campos, S V; Afonso, J E; Samano, M N; Pêgo-Fernandes, P M; Dolhnikoff, M; Teixeiraa, R H O B; Costa, A N

2017-05-01

Gastro-esophageal reflux disease (GERD) and broncho-aspiration (BA) are known to increase the risk for chronic lung allograft dysfunction (CLAD). However, specific lung injury mechanisms are not clearly known. The objective of the study was to describe histopathological findings in surveillance lung transbronchial biopsies that can be correlated with episodes of BA in the lung allograft. This retrospective analysis of surveillance transbronchial biopsies was performed in lung transplant recipients, with available data of broncho-alveolar fluid (cultures and cytology), lung function parameters, and esophageal functional tests. Were analyzed 11 patients, divided into 3 groups: (1) GERD group: 4 patients with GERD and CLAD diagnosis; (2) control group: 2 patients without GERD or CLAD; and (3) BA group: 5 patients with foreign material in lung biopsies. A histopathological pattern of neutrophilic bronchitis (NB) was present in 4 of 4 cases in the GERD group and in 1 of 5 cases in the BA group in 2 or more biopsy samples; culture samples were all negative; the 5 NB-positive patients developed CLAD and died (3/5) or needed re-transplantation (2/5). The other 3 patients in the BA group had GERD without NB or CLAD. Both patients in the control group had transient NB in biopsies with positive cultures but remained free of CLAD. Surveillance transbronchial biopsies may provide useful information other than the evaluation of acute cellular rejection and can help to identify high-risk patients for allograft dysfunction related to gastro-esophageal reflux. Copyright © 2017 Elsevier Inc. All rights reserved.

The truth about laser fiber diameters.

PubMed

Kronenberg, Peter; Traxer, Olivier

2014-12-01

To measure the various diameters of laser fibers from various manufacturers and compare them with the advertised diameter. Fourteen different unused laser fibers from 6 leading manufacturers with advertised diameters of 200, 270, 272, 273, 365, and 400 μm were measured by light microscopy. The outer diameter (including the fiber coating, cladding, and core), cladding diameter (including the cladding and the fiber core), and core diameter were measured. Industry representatives of the manufacturers were interviewed about the diameter of their fibers. For all fibers, the outer and cladding diameters differed significantly from the advertised diameter (P <.00001). The outer diameter, which is of most practical relevance for urologists, exhibited a median increase of 87.3% (range, 50.7%-116.7%). The outer, cladding, and core diameters of fibers with equivalent advertised diameters differed by up to 180, 100, and 78 μm, respectively. Some 200-μm fibers had larger outer diameters than the 270- to 273-μm fibers. All packaging material and all laser fibers lacked clear and precise fiber diameter information labels. Of 12 representatives interviewed, 8, 3, and 1 considered the advertised diameter to be the outer, the cladding, and the core diameter, respectively. Representatives within the same company frequently gave different answers. This study suggests that, at present, there is a lack of uniformity between laser fiber manufacturers, and most of the information conveyed to urologists regarding laser fiber diameter may be incorrect. Because fibers larger than the advertised laser fibers are known to influence key interventional parameters, this misinformation can have surgical repercussions. Copyright © 2014 Elsevier Inc. All rights reserved.

FY-16 Technology Gap Study Technical Report: Analysis of Undissolved Anode Materials of Mark-IV Electrorefiner

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

Yoo, Tae-Sic; Vaden, DeeEarl; Westphal, Brian Robert

2016-01-01

The Experimental Breeder Reactor II (EBR-II) is a sodium cooled fast reactor developed at Argonne National Laboratory (ANL). The used fuels from the EBR-II are currently being treated in the Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL). The Mark IV (Mk-IV) electrorefiner (ER) is a unit process in the FCF, which is primarily assigned to treating the used driver fuels. The stainless steel anode baskets hold the chopped spent driver fuel segments. During electrorefining, the anode baskets are immersed into the electrolyte and the used fuel is dissolved electrochemically. Perforated sides and bottoms allow the flow ofmore » the electrolyte into and out of the anode baskets. The steel cathode is also immersed into the electrolyte and collects the reduced products. The active metal contents in the used fuel (e.g., Cs, Sr, lanthanides, Pu, etc.) reacts with uranium cations in the electrolyte and progressively reports to the electrolyte. Noble metals are mostly retained in the cladding hulls. Varying quantities of zirconium are retained in the cladding hulls depending on the operational conditions of the Mk-IV ER. The undissolved anode materials are removed from the anode baskets and stored for subsequent metal waste form processing. These undissolved materials typically include undissolved fuels, stainless steel cladding, and adhering electrolyte. A couple of hulls are retrieved for chemical analysis and used for estimating the composition of the entire undissolved anode materials. The mass balance attempt based on this practice of estimating the undissolved anode materials has been a challenge due to inherently high sampling errors associated with heterogeneous undissolved material compositions. Responding to the prescribed challenge, this report investigates chemical analysis data as a whole and finds noticeable trends in the compositions of undissolved anode material samples with respect to the mass of the whole undissolved anode materials. Based upon this discovery, an empirical model is proposed.« less

Estimation of photonic band gap in the hollow core cylindrical multilayer structure

NASA Astrophysics Data System (ADS)

Chourasia, Ritesh Kumar; Singh, Vivek

2018-04-01

The propagation characteristic of two hollow core cylindrical multilayer structures having high and low refractive index contrast of cladding regions have been studied and compared at two design wavelengths i.e. 1550 nm and 632.8 nm. With the help of transfer matrix method a relation between the incoming light wave and outgoing light wave has been developed using the boundary matching technique. In high refractive index contrast, small numbers of layers are sufficient to provide perfect band gap in both design wavelengths. The spectral position and width of band gap is highly depending on the optical path of incident light in all considered cases. For sensing application, the sensitivity of waveguide can be obtained either by monitoring the width of photonic band gap or by monitoring the spectral shift of photonic band gap. Change in the width of photonic band gap with the core refractive index is larger in high refractive index contrast of cladding materials. However, in the case of monitoring the spectral shift of band gap, the obtained sensitivity is large for low refractive index contrast of cladding materials and further it increases with increase of design wavelength.

Tensile properties and flow behavior analysis of modified 9Cr-1Mo steel clad tube material

NASA Astrophysics Data System (ADS)

Singh, Kanwarjeet; Latha, S.; Nandagopal, M.; Mathew, M. D.; Laha, K.; Jayakumar, T.

2014-11-01

The tensile properties and flow behavior of modified 9Cr-1Mo steel clad tube have been investigated in the framework of various constitutive equations for a wide range of temperatures (300-923 K) and strain rates (3 × 10-3 s-1, 3 × 10-4 s-1 and 3 × 10-5 s-1). The tensile flow behavior of modified 9Cr-1Mo steel clad tube was most accurately described by Voce equation. The variation of instantaneous work hardening rate (θ = dσ/dε) and σθ with stress (σ) indicated two stage behavior characterized by rapid decrease at low stresses (transient stage) followed by a gradual decrease in high stresses (Stage III). The variation of work hardening parameters and work hardening rate in terms of θ vs. σ and σθ vs. σ with temperature exhibited three distinct regimes. Rapid decrease in flow stress and work hardening parameters and rapid shift of θ vs. σ and σθ vs. σ towards low stresses with increase in temperature indicated dynamic recovery at high temperatures. Tensile properties of the material have been best predicted from Voce equation.

Material selection for accident tolerant fuel cladding

DOE PAGES

Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; ...

2015-09-14

Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ≥1200°C for short (≤4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti 2AlC form a protective alumina scale in steam. Therefore, commercial Ti 2AlC that is not single phase, formed a much thicker oxide at 1200°Cmore » in steam and significant TiO 2, and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich α’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated.« less

Advanced Metalworking Solutions for Naval Systems That Go in Harm’s Way.

DTIC Science & Technology

2011-01-01

Cox, Titanium Fabrication Corporation, MMC, NSWCCD, ABS, and NMC. Navy Metalworking Center • Advanced Metallic Materials NMC has a successful record...Current efforts involve titanium , high-strength steel, and other alternate materials. 4 ADVANcED METALLic MATEriALS A cost-effective manufacturing solution...Manufacturing and Sustainment Technologies (iMAST). Improved shaft cladding materials and processes, which will increase the life of the main propulsion

Computational Analysis of Splash Occurring in the Deposition Process in Annular-Mist Flow

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

Xie, Heng; Koshizuka, Seiichi; Oka, Yoshiaki

2004-07-01

The deposition process of a single droplet on the film is numerically simulated by the Moving Particle Semi-implicit (MPS) method to analyze the possibility and effect of splash occurring in the deposition process in BWR condition. The model accounts for the presence of inertial, gravitation, viscous and surface tension and is validated by comparison with experiment results. A simple one-dimensional mixture model is developed to calculate the necessary parameters for the simulation of deposition in BWR condition. The deposition process of a single droplet in BWR condition is simulated. The effect of impact angle of droplet and the velocity ofmore » liquid film are analyzed. A film buffer model is developed to fit the simulation results of critical value for splash. A correlation of critical Weber number for splash in BWR condition is obtained and used to analyze the effect of splash. It is found that the splash play important role in the deposition and re-entrainment process in high quality condition in BWR. The mass fraction of re-entrainment caused by splash in different quality condition is also calculated. (authors)« less

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

J.C. Ryman

This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-00006 REV 01. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operatingmore » conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 7.28). The performance of the calculation and development of this document are carried out in accordance with AP-3.124, ''Design Calculation and Analyses'' (Ref. 7.29).« less

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

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

Gorzel, A.

2006-07-01

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a Germanmore » BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)« less

Self-sensing E-glass-fiber-reinforced composites

NASA Astrophysics Data System (ADS)

Brooks, David; Hayes, Simon A.; Khan, N. A.; Zolfaghar, K.; Fernando, Gerard F.

1997-06-01

Conventional E-glass fibers were surface treated to enable them to act as light guides for short distances. The reinforcing fiber light guides were embedded in glass fiber reinforced epoxy prepregs and processed into composites. The resultant composite was termed the self-sensing composite as any damage to these fibers or its interface would result in the attenuation of the transmitted light. Epoxy, silicone, fluoropolymer and sol-gel derived cladding materials were evaluated as potential cladding materials. RFLGs with a silicone coating was found to give the best light transmission. The self-sensing fibers were capable of detecting a 0.5 J direct impact. The feasibility of using the RFLGs for impact damage location was also demonstrated successfully as bleeding-light could be seen in the vicinity of the impact.

Recent GE BWR fuel experience and design evolution

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

Wood, J.E.; Potts, G.A.; Proebstle, R.A.

1992-01-01

Reliable fuel operation is essential to the safe, reliable, and economic power production by today's commercial nuclear reactors. GE Nuclear Energy is committed to maximize fuel reliability through the progressive development of improved fuel design features and dedication to provide the maximum quality of the design features and dedication to provide the maximum quality of the design, fabrication, and operation of GE BWR fuel. Over the last 35 years, GE has designed, fabricated, and placed in operation over 82,000 BWR fuel bundles containing over 5 million fuel rods. This experience includes successful commercial reactor operation of fuel assemblies to greatermore » than 45000 MWd/MTU bundle average exposure. This paper reports that this extensive experience base has enabled clear identification and characterization of the active failure mechanisms. With this failure mechanism characterization, mitigating actions have been developed and implemented by GE to provide the highest reliability BWR fuel bundles possible.« less

Numerical Modeling of Tube Forming by HPTR Cold Pilgering Process

NASA Astrophysics Data System (ADS)

Sornin, D.; Pachón-Rodríguez, E. A.; Vanegas-Márquez, E.; Mocellin, K.; Logé, R.

2016-09-01

For new fast-neutron sodium-cooled Generation IV nuclear reactors, the candidate cladding materials for the very strong burn-up are ferritic and martensitic oxide dispersion strengthened grades. Classically, the cladding tube is cold formed by a sequence of cold pilger milling passes with intermediate heat treatments. This process acts upon the geometry and the microstructure of the tubes. Consequently, crystallographic texture, grain sizes and morphologies, and tube integrity are highly dependent on the pilgering parameters. In order to optimize the resulting mechanical properties of cold-rolled cladding tubes, it is essential to have a thorough understanding of the pilgering process. Finite Element Method (FEM) models are used for the numerical predictions of this task; however, the accuracy of the numerical predictions depends not only on the type of constitutive laws but also on the quality of the material parameters identification. Therefore, a Chaboche-type law which parameters have been identified on experimental observation of the mechanical behavior of the material is used here. As a complete three-dimensional FEM mechanical analysis of the high-precision tube rolling (HPTR) cold pilgering of tubes could be very expensive, only the evolution of geometry and deformation is addressed in this work. The computed geometry is compared to the experimental one. It is shown that the evolution of the geometry and deformation is not homogeneous over the circumference. Moreover, it is exposed that the strain is nonhomogeneous in the radial, tangential, and axial directions. Finally, it is seen that the dominant deformation mode of a material point evolves during HPTR cold pilgering forming.

76 FR 72441 - Certain Steel Wheels From China; Scheduling of the Final Phase of Countervailing Duty and...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-23

....S.C. 1673d(b)) to determine whether an industry in the United States is materially injured or threatened with material injury, or the establishment of an industry in the United States is materially... carbon and/or alloy composition and clad wheels, discs, and rims when carbon or alloy steel represents...

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale FeCrAl Modeling

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

Gamble, K. A.; Hales, J. D.; Zhang, Y.

Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. Thismore » allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to utilize lower length scale approaches (e.g., density functional theory, cluster dynamics, rate theory, phase field, and Visco-Plastic- Self-Consistent (VPSC)) to develop more physically informed models at the engineering scale for use in the BISON [9] fuel performance code.« less

A study on wear resistance and microcrack of the Ti 3Al/TiAl + TiC ceramic layer deposited by laser cladding on Ti-6Al-4V alloy

NASA Astrophysics Data System (ADS)

Li, Jianing; Chen, Chuanzhong; Squartini, Tiziano; He, Qingshan

2010-12-01

Laser cladding of the Al + TiC alloy powder on Ti-6Al-4V alloy can form the Ti 3Al/TiAl + TiC ceramic layer. In this study, TiC particle-dispersed Ti 3Al/TiAl matrix ceramic layer on the Ti-6Al-4V alloy by laser cladding has been researched by means of X-ray diffraction, scanning electron microscope, electron probe micro-analyzer, energy dispersive spectrometer. The main difference from the earlier reports is that Ti 3Al/TiAl has been chosen as the matrix of the composite coating. The wear resistance of the Al + 30 wt.% TiC and the Al + 40 wt.% TiC cladding layer was approximately 2 times greater than that of the Ti-6Al-4V substrate due to the reinforcement of the Ti 3Al/TiAl + TiC hard phases. However, when the TiC mass percent was above 40 wt.%, the thermal stress value was greater than the materials yield strength limit in the ceramic layer, the microcrack was present and its wear resistance decreased.

Laser cladding of stainless steel with a copper-silver alloy to generate surfaces of high antimicrobial activity

NASA Astrophysics Data System (ADS)

Hans, Michael; Támara, Juan Carlos; Mathews, Salima; Bax, Benjamin; Hegetschweiler, Andreas; Kautenburger, Ralf; Solioz, Marc; Mücklich, Frank

2014-11-01

Copper and silver are used as antimicrobial agents in the healthcare sector in an effort to curb infections caused by bacteria resistant to multiple antibiotics. While the bactericidal potential of copper and silver alone are well documented, not much is known about the antimicrobial properties of copper-silver alloys. This study focuses on the antibacterial activity and material aspects of a copper-silver model alloy with 10 wt% Ag. The alloy was generated as a coating with controlled intermixing of copper and silver on stainless steel by a laser cladding process. The microstructure of the clad was found to be two-phased and in thermal equilibrium with minor Cu2O inclusions. Ion release and killing of Escherichia coli under wet conditions were assessed with the alloy, pure silver, pure copper and stainless steel. It was found that the copper-silver alloy, compared to the pure elements, exhibited enhanced killing of E. coli, which correlated with an up to 28-fold increased release of copper ions. The results show that laser cladding with copper and silver allows the generation of surfaces with enhanced antimicrobial properties. The process is particularly attractive since it can be applied to existing surfaces.

Surface modification of AISI H13 tool steel by laser cladding with NiTi powder

NASA Astrophysics Data System (ADS)

Norhafzan, B.; Aqida, S. N.; Chikarakara, E.; Brabazon, D.

2016-04-01

This paper presents laser cladding of NiTi powder on AISI H13 tool steel surface for surface properties enhancement. The cladding process was conducted using Rofin DC-015 diffusion-cooled CO2 laser system with wavelength of 10.6 µm. NiTi powder was pre-placed on H13 tool steel surface. The laser beam was focused with a spot size of 90 µm on the sample surface. Laser parameters were set to 1515 and 1138 W peak power, 18 and 24 % duty cycle and 2300-3500 Hz laser pulse repetition frequency. Hardness properties of the modified layer were characterized by Wilson Hardness tester. Metallographic study and chemical composition were conducted using field emission scanning electron microscope and energy-dispersive X-ray spectrometer (EDXS) analysis. Results showed that hardness of NiTi clad layer increased three times that of the substrate material. The EDXS analysis detected NiTi phase presence in the modified layer up to 9.8 wt%. The metallographic study shows high metallurgical bonding between substrate and modified layer. These findings are significant to both increased hardness and erosion resistance of high-wear-resistant components and elongating their lifetime.

Multifrequency Eddy Current Inspection of Corrosion in Clad Aluminum Riveted Lap Joints and Its Effect on Fatigue Life

NASA Astrophysics Data System (ADS)

Okafor, A. C.; Natarajan, S.

2007-03-01

Aging aircraft are prone to corrosion damage and fatigue cracks in riveted lap joints of fuselage skin panels. This can cause catastrophic failure if not detected and repaired. Hence detection of corrosion damage and monitoring its effect on structural integrity are essential. This paper presents multifrequency eddy current (EC) inspection of corrosion damage and machined material loss defect in clad A1 2024-T3 riveted lap joints and its effect on fatigue life. Results of eddy current inspection, corrosion product removal and fatigue testing are presented.

Structure and Performance of Epoxy Resin Cladded Graphite Used as Anode

NASA Astrophysics Data System (ADS)

Zhou, Zhentao; Li, Haijun

This paper is concerning to prepare modified natural graphite which is low-cost and advanced materials used as lithium ion battery anode using the way of cladding natural graphite with epoxy resin. The results shows that the specific capacity and circular performance of the modified natural graphite, which is prepared in the range of 600°C and 1000°C, have been apparently improved compare with the not-modified natural graphite. The first reversible capacity of the modified natural graphite is 338mAh/g and maintain more than 330mAh/g after 20 charge/discharge circles.

Influence of depressed-index outer ring on evanescent tunneling loss in tapered double-cladding fibers.

PubMed

Chen, Nan-Kuang; Hsu, Kuei-Chu; Liaw, Shien-Kuei; Lai, Yinchieh; Chi, Sien

2008-08-01

A tapered fiber with a depressed-index outer ring is fabricated and dispersion engineered to generate a widely tunable (1250-1650 nm) fundamental-mode leakage loss with a high cutoff slope (-1.2 dB/nm) and a high attenuation for stop band (>50 dB) by modification of both waveguide and material dispersions. The higher cutoff slope is achieved with a larger cross angle between the two refractive index dispersion curves of the tapered fiber and surrounding optical liquids through the use of depressed-index outer ring structures in double-cladding fibers.

Stressed waveguides with tubular depressed-cladding inscribed in phosphate glasses by femtosecond hollow laser beams.

PubMed

Long, Xuewen; Bai, Jing; Zhao, Wei; Stoian, Razvan; Hui, Rongqing; Cheng, Guanghua

2012-08-01

We report on the single-step fabrication of stressed optical waveguides with tubular depressed-refractive-index cladding in phosphate glasses by the use of focused femtosecond hollow laser beams. Tubelike low index regions appear under direct exposure due to material rarefaction following expansion. Strained compacted zones emerged in domains neighboring the tubular track of lower refractive index, and waveguiding occurs mainly within the tube core fabricated by the engineered femtosecond laser beam. The refractive index profile of the optical waveguide was reconstructed from the measured transmitted near-field intensity.

Carbide fuel pin and capsule design for irradiations at thermionic temperatures

NASA Technical Reports Server (NTRS)

Siegel, B. L.; Slaby, J. G.; Mattson, W. F.; Dilanni, D. C.

1973-01-01

The design of a capsule assembly to evaluate tungsten-emitter - carbide-fuel combinations for thermionic fuel elements is presented. An inpile fuel pin evaluation program concerned with clad temperture, neutron spectrum, carbide fuel composition, fuel geometry,fuel density, and clad thickness is discussed. The capsule design was a compromise involving considerations between heat transfer, instrumentation, materials compatibility, and test location. Heat-transfer calculations were instrumental in determining the method of support of the fuel pin to minimize axial temperature variations. The capsule design was easily fabricable and utilized existing state-of-the-art experience from previous programs.

Pu-Zr alloy for high-temperature foil-type fuel

DOEpatents

McCuaig, Franklin D.

1977-01-01

A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron reflux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

Pu-ZR Alloy high-temperature activation-measurement foil

DOEpatents

McCuaig, Franklin D.

1977-08-02

A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron flux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

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

K. L. Davis; D. L. Knudson; J. L. Rempe

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status ofmore » INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.« less

Process for making a martensitic steel alloy fuel cladding product

DOEpatents

Johnson, Gerald D.; Lobsinger, Ralph J.; Hamilton, Margaret L.; Gelles, David S.

1990-01-01

This is a very narrowly defined martensitic steel alloy fuel cladding material for liquid metal cooled reactors, and a process for making such a martensitic steel alloy material. The alloy contains about 10.6 wt. % chromium, about 1.5 wt. % molybdenum, about 0.85 wt. % manganese, about 0.2 wt. % niobium, about 0.37 wt. % silicon, about 0.2 wt. % carbon, about 0.2 wt. % vanadium, 0.05 maximum wt. % nickel, about 0.015 wt. % nitrogen, about 0.015 wt. % sulfur, about 0.05 wt. % copper, about 0.007 wt. % boron, about 0.007 wt. % phosphorous, and with the remainder being essentially iron. The process utilizes preparing such an alloy and homogenizing said alloy at about 1000.degree. C. for 16 hours; annealing said homogenized alloy at 1150.degree. C. for 15 minutes; and tempering said annealed alloy at 700.degree. C. for 2 hours. The material exhibits good high temperature strength (especially long stress rupture life) at elevated temperature (500.degree.-760.degree. C.).

A high-throughput investigation of Fe-Cr-Al as a novel high-temperature coating for nuclear cladding materials.

PubMed

Bunn, Jonathan Kenneth; Fang, Randy L; Albing, Mark R; Mehta, Apurva; Kramer, Matthew J; Besser, Matthew F; Hattrick-Simpers, Jason R

2015-07-10

High-temperature alloy coatings that can resist oxidation are urgently needed as nuclear cladding materials to mitigate the danger of hydrogen explosions during meltdown. Here we apply a combination of computationally guided materials synthesis, high-throughput structural characterization and data analysis tools to investigate the feasibility of coatings from the Fe–Cr–Al alloy system. Composition-spread samples were synthesized to cover the region of the phase diagram previous bulk studies have identified as forming protective oxides. The metallurgical and oxide phase evolution were studied via in situ synchrotron glancing incidence x-ray diffraction at temperatures up to 690 K. A composition region with an Al concentration greater than 3.08 at%, and between 20.0 at% and 32.9 at% Cr showed the least overall oxide growth. Subsequently, a series of samples were deposited on stubs and their oxidation behavior at 1373 K was observed. The continued presence of a passivating oxide was confirmed in this region over a period of 6 h.

Characterization of microstructure and property evolution in advanced cladding and duct: Materials exposed to high dose and elevated temperature

DOE PAGES

Allen, Todd R.; Kaoumi, Djamel; Wharry, Janelle P.; ...

2015-05-20

Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. Here, the studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocationmore » networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Lastly, transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.« less

Analysis of Operating Strategies Using Different Target Designs For 238Pu Production

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

Thomas, Tomcy; Sherman, Steven R; Sawhney, Dr. Rapinder

2017-01-01

An engineering effort is underway to re-establish capability to produce 238Pu oxide at the kilogram scale in the United States. A multi-step batch process is being developed to produce this important material. Recently, a portion of this process was studied using discrete-event simulation tools to determine whether the conceptual process might achieve its yearly production goal. The study showed the conceptual process can meet the yearly production goal under some circumstances, but process improvements would be needed to ensure greater likelihood of success. This study extends the work performed previously by examining the effects of changing the reactor target designmore » on the yearly process output. Two new reactor target configurations are considered an aluminum-clad reactor target containing 50% greater 237Np oxide content than the original target, and a zirconium alloy-clad target using no aluminum. The results indicate that use of the new aluminum-clad target configuration may allow the process to achieve the same yearly production goal in less time using fewer targets. If the zirconium alloy-clad target is used, then even fewer targets would be needed to reach the production goal, but some process changes would be required to handle the zirconium cladding. The number of days needed to process a target batch to completion, and the steady state 238Pu oxide production rate, for each configuration are compared to the results from the initial simulation study.« less

Study on Pot Forming of Induction Heater Type Rice Cookers by Forging Cast Process

NASA Astrophysics Data System (ADS)

Ohnishi, Masayuki; Yamaguchi, Mitsugi; Ohashi, Osamu

This paper describes a study result on pot fabrication by the forging cast process of stainless steel with aluminum. Rice cooked with the new bowl-shaped pot for the induction heater type rice cookers is better tasting than rice cooked with the conventional cylindrical one, due to the achievement of better heat conduction and convection. The conventional pot is made of the clad sheet, consisting of stainless steel and aluminum. However, it is rather difficult to form a bowl shape from the clad sheet, primarily due to the problem of a material spring back. The fabrication of a new type of a pot was made possible by means of the adoption of a forging cast process instead of the clad sheet. In this process, iron powder is inserted between stainless steel and aluminum in order to alleviate the large difference on the coefficient of expansion between each material. It was made clear that the application of two kinds of iron particle, namely 10 μm size powder on the stainless steel side and 44 μm on the aluminum side, enables the joints to become strong enough. The joint strength of the new pot by this fabrication process was confirmed by the tests of the shear strength and the fatigue tests together with the stress analysis.

Enhanced control of electrochemical response in metallic materials in neural stimulation electrode applications

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

Watkins, K.G.; Steen, W.M.; Manna, I.

New means have been investigated for the production of electrode devices (stimulation electrodes) which could be implanted in the human body in order to control pain, activate paralysed limbs or provide electrode arrays for cochlear implants for the deaf or for the relief of tinitus. To achieve this ion implantation and laser materials processing techniques were employed. Ir was ion implanted in Ti-6Al-4V alloy and the surface subsequently enriched in the noble metal by dissolution in sulphuric acid. For laser materials processing techniques, investigation has been carried out on the laser cladding and laser alloying of Ir in Ti wire.more » A particular aim has been the determination of conditions required for the formation of a two phase Ir, Ir-rich, and Ti-rich microstructure which would enable subsequent removal of the non-noble phase to leave a highly porous noble metal with large real surface area and hence improved charge carrying capacity compared with conventional non porous electrodes. Evaluation of the materials produced has been carried out using repetitive cyclic voltammetry, amongst other techniques. For laser alloyed Ir on Ti wire, it has been found that differences in the melting point and density of the materials makes control of the cladding or alloying process difficult. Investigation of laser process parameters for the control of alloying and cladding in this system was carried out and a set of conditions for the successful production of two phase Ir-rich and Ti-rich components in a coating layer with strong metallurgical bonding to the Ti alloy substrate was derived. The laser processed material displays excellent potential for further development in providing stimulation electrodes with the current carrying capacity of Ir but in a form which is malleable and hence capable of formation into smaller electrodes with improved spatial resolution compared with presently employed electrodes.« less

Study of self-ion irradiated nanostructured ferritic alloy (NFA) and silicon carbide-nanostructured ferritic alloy (SiC-NFA) cladding materials

NASA Astrophysics Data System (ADS)

Ning, Kaijie; Bai, Xianming; Lu, Kathy

2018-07-01

Silicon carbide-nanostructured ferritic alloy (SiC-NFA) materials are expected to have the beneficial properties of each component for advanced nuclear claddings. Fabrication of pure NFA (0 vol% SiC-100 vol% NFA) and SiC-NFAs (2.5 vol% SiC-97.5 vol% NFA, 5 vol% SiC-95 vol% NFA) has been reported in our previous work. This paper is focused on the study of radiation damage in these materials under 5 MeV Fe++ ion irradiation with a dose up to ∼264 dpa. It is found that the material surfaces are damaged to high roughness with irregularly shaped ripples, which can be explained by the Bradley-Harper (B-H) model. The NFA matrix shows ion irradiation induced defect clusters and small dislocation loops, while the crystalline structure is maintained. Reaction products of Fe3Si and Cr23C6 are identified in the SiC-NFA materials, with the former having a partially crystalline structure but the latter having a fully amorphous structure upon irradiation. The different radiation damage behaviors of NFA, Fe3Si, and Cr23C6 are explained using the defect reaction rate theory.

Ceramics reinforced metal base composite coatings produced by CO II laser cladding

NASA Astrophysics Data System (ADS)

Yang, Xichen; Wang, Yu; Yang, Nan

2008-03-01

Due to the excellent performance in high strength, anti-temperature and anti-wear, ceramics reinforced metal base composite material was used in some important fields of aircraft, aerospace, automobile and defense. The traditional bulk metal base composite materials are the expensive cost, which is limited in its industrial application. Development of laser coating of ceramics reinforced metal base composite is very interesting in economy. This paper is focused on three laser cladding ceramics coatings of SiC particle /Al matrix , Al IIO 3 powder/ Al matrix and WC + Co/mild steel matrix. Powder particle sizes are of 10-60μm. Chemical contents of aluminum matrix are of 3.8-4.0% Cu, 1.2-1.8% Mg, 0.3-0.99% Mn and balance Al. 5KW CO II laser, 5 axes CNC table, JKF-6 type powder feeder and co-axis feeder nozzle are used in laser cladding. Microstructure and performance of laser composite coatings have been respectively examined with OM,SEM and X-ray diffraction. Its results are as follows : Microstructures of 3C-,6H- and 5H- SiC particles + Al + Al 4SiC 4 + Si in SiC/Al composite, hexagonal α-Al IIO 3 + cubic γ-Al IIO 3 + f.c.c Al in Al IIO 3 powder/ Al composite and original WC particles + separated WC particles + eutectic WC + γ-Co solid solution + W IIC particles in WC + Co/steel coatings are respectively recognized. New microstructures of 5H-SiC in SiC/Al composite, cubic γ-Al IIO 3 in Al IIO 3 composite and W IIC in WC + Co/ steel composite by laser cladding have been respectively observed.

Derivative effect of laser cladding on interface stability of YSZ@Ni coating on GH4169 alloy: An experimental and theoretical study

NASA Astrophysics Data System (ADS)

Zheng, Haizhong; Li, Bingtian; Tan, Yong; Li, Guifa; Shu, Xiaoyong; Peng, Ping

2018-01-01

Yttria-stabilized zirconia YSZ@Ni core-shell nanoparticles were used to prepare a thermal barrier coating (TBC) on a GH4169 alloy by laser cladding. Microstructural analysis showed that the TBC was composed of two parts: a ceramic and a bonding layer. In places where the ZrO2/Al2O3 eutectic structure was present in the ceramic layer, the Ni atoms diffused into the bonding layer, as confirmed by energy-dispersive X-ray spectroscopy (EDS). The derivative effect of laser cladding results in the original YSZ@Ni core-shell nanoparticles being translated into the Al2O3 crystal, activating the YSZ. The mechanism of ceramic/metal interface cohesion was studied in depth via first-principles and molecular dynamics simulation. The results show that the trend in the diffusion coefficients of Ni, Fe, Al, and Ti is DNi > DFe > DTi > DAl in the melting or solidification process of the material. The enthalpy of formation for Al2O3 is less than that of TiO2, resulting in a thermally grown oxide (TGO) Al2O3 phase transformation. With regard to the electronic structure, the trend in Mulliken population is QO-Ni > QZr-O > QO-Al. Although the bonding is slightly weakened between ZrO2/Al2O3 (QZr-O = 0.158 < QO-Ni = 0.220) compared to that in ZrO2/Ni, TGO Al2O3 can improve the oxidation resistance of the metal matrix. Thus, by comparing the connective and diffusive processes, our findings lay the groundwork for detailed and comprehensive studies of the laser cladding process for the production of composite materials.

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

Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as thesemore » installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.« less

Window Frame Types | Efficient Windows Collaborative

Science.gov Websites

metal frames. Metal Frames Metal Frame with Thermal Break Non-metal Frames Non-metal There is a variety of non-metal framing materials for windows including, wood, wood with metal/vinyl cladding, vinyl disadvantages. Non-metal Frames Non-metal Frame, Thermally Improved Does frame material type matter? The

Biotronics

DTIC Science & Technology

2012-11-01

BLOCKING LAYER IN ORGANIC LIGHT EMITTING DIODES ............................70 2.3.1 Materials Used for the Fabrication of BioLEDs...optical losses. Using a lower molecular weight DNA-based biopolymer as the top and bottom cladding layers in an NLO polymer EO modulator, we were able...application, these new biopolymer -based materials have been used for many types of electronic and photonic applications. Even with growing research

Thermal-Hydraulic Results for the Boiling Water Reactor Dry Cask Simulator.

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

Durbin, Samuel; Lindgren, Eric R.

The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internalmore » convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both aboveground and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of this investigation was to produce validation-quality data that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. These cladding temperatures are critical to evaluate cladding integrity throughout the storage cycle. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplified interpretation of results. Two different arrangements of ducting were used to mimic conditions for aboveground and belowground storage configurations for vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured throughout the test assembly. The induced air mass flow rate was measured for both the aboveground and belowground configurations. In addition, the impact of cross-wind conditions on the belowground configuration was quantified. Over 40 unique data sets were collected and analyzed for these efforts. Fourteen data sets for the aboveground configuration were recorded for powers and internal pressures ranging from 0.5 to 5.0 kW and 0.3 to 800 kPa absolute, respectively. Similarly, fourteen data sets were logged for the belowground configuration starting at ambient conditions and concluding with thermal-hydraulic steady state. Over thirteen tests were conducted using a custom-built wind machine. The results documented in this report highlight a small, but representative, subset of the available data from this test series. This addition to the dry cask experimental database signifies a substantial addition of first-of-a-kind, high-fidelity transient and steady-state thermal-hydraulic data sets suitable for CFD model validation.« less

78 FR 18375 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-26

... Pike, Rockville, Maryland. Thursday, April 11, 2013, Conference Room T2-B1, 11545 Rockville Pike..., ``Westinghouse BWR ECCS Evaluation Model: Supplement 5--Application to the ABWR,'' Revision 0 (Open/Closed)--The...-17116-P, ``Westinghouse BWR Emergency Core Coolant System (ECCS) Evaluation Model: Supplement 5,'' and...

A Multi-Stage Wear Model for Grid-to-Rod Fretting of Nuclear Fuel Rods

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

Blau, Peter Julian

The wear of fuel rod cladding against the supporting structures in the cores of pressurized water nuclear reactors (PWRs) is an important and potentially costly tribological issue. Grid-to-rod fretting (GTRF), as it is known, involves not only time-varying contact conditions, but also elevated temperatures, flowing hot water, aqueous tribo-corrosion, and the embrittling effects of neutron fluences. The multi-stage, closed-form analytical model described in this paper relies on published out-of-reactor wear and corrosion data and a set of simplifying assumptions to portray the conversion of frictional work into wear depth. The cladding material of interest is a zirconium-based alloy called Zircaloy-4,more » and the grid support is made of a harder and more wear-resistant material. Focus is on the wear of the cladding. The model involves an incubation stage, a surface oxide wear stage, and a base alloy wear stage. The wear coefficient, which is a measure of the efficiency of conversion of frictional work into wear damage, can change to reflect the evolving metallurgical condition of the alloy. Wear coefficients for Zircaloy-4 and for a polyphase zirconia layer were back-calculated for a range of times required to wear to a critical depth. Inputs for the model, like the friction coefficient, are taken from the tribology literature in lieu of in-reactor tribological data. Concepts of classical fretting were used as a basis, but are modified to enable the model to accommodate the complexities of the PWR environment. Factors like grid spring relaxation, pre-oxidation of the cladding, multiple oxide phases, gap formation, impact, and hydrogen embrittlement are part of the problem definition but uncertainties in their relative roles limits the ability to validate the model. Sample calculations of wear depth versus time in the cladding illustrate how GTRF wear might occur in a discontinuous fashion during months-long reactor operating cycles. A means to account for grid/rod gaps and repetitive impact effects on GTRF wear is proposed« less

Laser cladding of bioactive glass coatings.

PubMed

Comesaña, R; Quintero, F; Lusquiños, F; Pascual, M J; Boutinguiza, M; Durán, A; Pou, J

2010-03-01

Laser cladding by powder injection has been used to produce bioactive glass coatings on titanium alloy (Ti6Al4V) substrates. Bioactive glass compositions alternative to 45S5 Bioglass were demonstrated to exhibit a gradual wetting angle-temperature evolution and therefore a more homogeneous deposition of the coating over the substrate was achieved. Among the different compositions studied, the S520 bioactive glass showed smoother wetting angle-temperature behavior and was successfully used as precursor material to produce bioactive coatings. Coatings processed using a Nd:YAG laser presented calcium silicate crystallization at the surface, with a uniform composition along the coating cross-section, and no significant dilution of the titanium alloy was observed. These coatings maintain similar bioactivity to that of the precursor material as demonstrated by immersion in simulated body fluid. Copyright 2009 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

High Resolution Neutron Radiography and Tomography of Hydrided Zircaloy-4 Cladding Materials

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

Smith, Tyler S; Bilheux, Hassina Z; Ray, Holly B

2015-01-01

Neutron radiography for hydrogen analysis was performed with several Zircaloy-4 cladding samples with controlled hydrogen concentrations up to 1100 ppm. Hydrogen charging was performed in a process tube that was heated to facilitate hydrogen absorption by the metal. A correlation between the hydrogen concentration in the hydrided tubes and the neutron intensity was established, by which hydrogen content can be determined precisely in a small area (55 m x 55 m). Radiography analysis was also performed to evaluate the heating rate and its correlation with the hydrogen distribution through hydrided materials. In addition to radiography analysis, tomography experiments were performedmore » on Zircaloy-4 tube samples to study the local hydrogen distribution. Through tomography analysis a 3D reconstruction of the tube was evaluated in which an uneven hydrogen distribution in the circumferential direction can be observed.« less

Toward single-mode active crystal fibers for next-generation high-power fiber devices.

PubMed

Lai, Chien-Chih; Gao, Wan-Ting; Nguyen, Duc Huy; Ma, Yuan-Ron; Cheng, Nai-Chia; Wang, Shih-Chang; Tjiu, Jeng-Wei; Huang, Chun-Ming

2014-08-27

We report what we believe to be the first demonstration of a facile approach with controlled geometry for the production of crystal-core ceramic-clad hybrid fibers for scaling fiber devices to high average powers. The process consists of dip coating a solution of polycrystalline alumina onto a high-crystallinity 40-μm-diameter Ti:sapphire single-crystalline core followed by thermal treatments. Comparison of the measured refractive index with high-resolution transmission electron microscopy reveals that a Ca/Si-rich intragranular layer is precipitated at grain boundaries by impurity segregation and liquid-phase formation due to the relief of misfit strain energy in the Al2O3 matrix, slightly perturbing the refractive index and hence the optical properties. Additionally, electron backscatter diffractions supply further evidence that the Ti:sapphire single-crystalline core provides the template for growth into a sacrificial polycrystalline cladding, bringing the core and cladding into a direct bond. The thus-prepared doped crystal core with the undoped crystal cladding was achieved through the abnormal grain-growth process. The presented results provide a general guideline both for controlling crystal growth and for the performance of hybrid materials and provides insights into how one might design single-mode high-power crystal fiber devices.

Process development and fabrication of space station type aluminum-clad graphite epoxy struts

NASA Technical Reports Server (NTRS)

Ring, L. R.

1990-01-01

The manufacture of aluminum-clad graphite epoxy struts, designed for application to the Space Station truss structure, is described. The strut requirements are identified, and the strut material selection rationale is discussed. The manufacturing procedure is described, and shop documents describing the details are included. Dry graphite fiber, Pitch-75, is pulled between two concentric aluminum tubes. Epoxy resin is then injected and cured. After reduction of the aluminum wall thickness by chemical milling the end fittings are bonded on the tubes. A discussion of the characteristics of the manufactured struts, i.e., geometry, weight, and any anomalies of the individual struts is included.

Model of depositing layer on cylindrical surface produced by induction-assisted laser cladding process

NASA Astrophysics Data System (ADS)

Kotlan, Václav; Hamar, Roman; Pánek, David; Doležel, Ivo

2017-12-01

A model of hybrid cladding on a cylindrical surface is built and numerically solved. Heating of both substrate and the powder material to be deposited on its surface is realized by laser beam and preheating inductor. The task represents a hard-coupled electromagnetic-thermal problem with time-varying geometry. Two specific algorithms are developed to incorporate this effect into the model, driven by local distribution of temperature and its gradients. The algorithms are implemented into the COMSOL Multiphysics 5.2 code that is used for numerical computations of the task. The methodology is illustrated with a typical example whose results are discussed.

Development of thin semi-rigid coaxial cables as low-pass filter using bilayer structure in center conductors

NASA Astrophysics Data System (ADS)

Kushino, Akihiro; Yamamoto, Yusei; Okuyama, Tetsuya; Kasai, Soichi

We have developed and evaluated thin semi-rigid coaxial cables as the noise filter for readout in low temperature experiments. The cables reported have 0.86 mm outer diameters consisting of seamless outer conductor, polytetrafluoroethylene (PTFE) dielectric, and center conductor made of superconducting niobium-titanium (NbTi). Each center conductor has surficial cladding made of normal conductor in different thickness. We had reported that we can adjust attenuation magnitude and cut-off frequency of the semi-rigid cable in the range about 100 500 MHz by controlling cable length and/or thickness of cladding. We newly manufactured this type of low-pass filter cables using stainless-steel (SUS304) as the material for cladding which has higher electrical resistivity than that of cupro-nickel (CuNi). It enables high filtering efficiency, i.e. large attenuation at the same frequency, compared to those made of conventional CuNi-based low-pass-filter cables.

Chemical vapor deposition of Mo tubes for fuel cladding applications

DOE PAGES

Beaux, Miles F.; Vodnik, Douglas R.; Peterson, Reuben J.; ...

2018-01-31

In this study, chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wallmore » thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. Finally, a path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system.« less

Chemical vapor deposition of Mo tubes for fuel cladding applications

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

Beaux, Miles F.; Vodnik, Douglas R.; Peterson, Reuben J.

In this study, chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wallmore » thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. Finally, a path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system.« less

An elasto-plastic fracture mechanics based model for assessment of hydride embrittlement in zircaloy cladding tubes

NASA Astrophysics Data System (ADS)

Nilsson, Karl-Fredrik; Jakšić, Nikola; Vokál, Vratko

2010-01-01

This paper describes a finite element based fracture mechanics model to assess how hydrides affect the integrity of zircaloy cladding tubes. The hydrides are assumed to fracture at a low load whereas the propagation of the fractured hydrides in the matrix material and failure of the tube is controlled by non-linear fracture mechanics and plastic collapse of the ligaments between the hydrides. The paper quantifies the relative importance of hydride geometrical parameters such as size, orientation and location of individual hydrides and interaction between adjacent hydrides. The paper also presents analyses for some different and representative multi-hydride configurations. The model is adaptable to general and complex crack configurations and can therefore be used to assess realistic hydride configurations. The mechanism of cladding failure is by plastic collapse of ligaments between interacting fractured hydrides. The results show that the integrity can be drastically reduced when several radial hydrides form continuous patterns.

Hot Cell Installation and Demonstration of the Severe Accident Test Station

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

Linton, Kory D.; Burns, Zachary M.; Terrani, Kurt A.

A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examinemore » postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.« less

Hydrogen motion in Zircaloy-4 cladding during a LOCA transient

NASA Astrophysics Data System (ADS)

Elodie, T.; Jean, D.; Séverine, G.; M-Christine, B.; Michel, C.; Berger, P.; Martine, B.; Antoine, A.

2016-04-01

Hydrogen and oxygen are key elements influencing the embrittlement of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). The understanding of the mechanisms influencing the motion of these two chemical elements in the metal is required to fully describe the material embrittlement. High temperature steam oxidation tests were performed on pre-hydrided Zircaloy-4 samples with hydrogen contents ranging between 11 and 400 wppm prior to LOCA transient. Thanks to the use of both Electron Probe Micro-Analysis (EPMA) and Elastic Recoil Detection Analysis (μ-ERDA), the chemical elements partitioning has been systematically quantified inside the prior-β phase. Image analysis and metallographic examinations were combined to provide an average oxygen profile as well as hydrogen profile within the cladding thickness after LOCA transient. The measured hydrogen profile is far from homogeneous. Experimental distributions are compared to those predicted numerically using calculations derived from a finite difference thermo-diffusion code (DIFFOX) developed at IRSN.

Residual Stress Measurement and the Effect of Heat Treatment in Cladded Control Rod Drive Specimens

NASA Astrophysics Data System (ADS)

Bowman, Ashley; Kingston, Ed; Katsuyama, Jinya; Udagawa, Makoto; Onizawa, Kunio

This paper presents results of residual stress measurements and modelling within the cladding and J-groove weld of Control Rod Drive (CRD) specimens in the as-welded and Post Weld Heat Treated (PWHT) states. Knowledge of the residual stresses present in CRD nozzles is critical when modelling the fracture mechanics of failures of nuclear power plant components to dictate inspections intervals and optimise plant downtime. The specimens comprised of ferritic steel blocks with 309L stainless steel cladding and a single J-groove weld attaching the 304 stainless steel nozzles. Multiple measurements were made through the thickness of the specimens in order to give biaxial residual stress profiles through all the different fusion boundaries. The results show the effect of PWHT in reducing residual stresses both in the weld and ferritic material. The beneficial use of measurements is highlighted to provide confidence in the modelled results and prevent over conservatism in integrity calculations, costing unnecessary time and money.

Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi

2016-06-01

We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV) of a Boiling Water Reactor (BWR) by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au) and Nickel (Ni) at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

DOE PAGES

Besmann, T. M.; Yamamoto, Y.; Unocic, K. A.

2016-06-21

We assessed the thermochemical compatibility of potential replacement cladding materials for zirconium alloys in light water reactors. Considered were FeCrAl steel (similar to Kanthal APMT), Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included a FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam, and for hybrid SiC-composites possible interactions between the metallic layer and SiC. Additionally, experimental exposures of SiC-alloymore » reaction couples at 673K, 1073K, and 1273K for 168 h in an inert atmosphere were made and microanalysis performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material with expected protective oxides forming. Finally, the computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.« less

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

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

Jacobsen, George; Back, Christina

2015-10-30

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called themore » endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.« less

Laser Cladding of γ-TiAl Intermetallic Alloy on Titanium Alloy Substrates

NASA Astrophysics Data System (ADS)

Maliutina, Iuliia Nikolaevna; Si-Mohand, Hocine; Piolet, Romain; Missemer, Florent; Popelyukh, Albert Igorevich; Belousova, Natalya Sergeevna; Bertrand, Philippe

2016-01-01

The enhancement of titanium and titanium alloy's tribological properties is of major interest in many applications such as the aerospace and automotive industry. Therefore, the current research paper investigates the laser cladding of Ti48Al2Cr2Nb powder onto Ti6242 titanium alloy substrates. The work was carried out in two steps. First, the optimal deposition parameters were defined using the so-called "combined parameters," i.e., the specific energy E specific and powder density G. Thus, the results show that those combined parameters have a significant influence on the geometry, microstructure, and microhardness of titanium aluminide-formed tracks. Then, the formation of dense, homogeneous, and defect-free coatings based on optimal parameters has been investigated. Optical and scanning electron microscopy techniques as well as energy-dispersive spectroscopy and X-ray diffraction analyses have shown that a duplex structure consisting of γ-TiAl and α 2-Ti3Al phases was obtained in the coatings during laser cladding. Moreover, it was shown that produced coatings exhibit higher values of microhardness (477 ± 9 Hv0.3) and wear resistance (average friction coefficient is 0.31 and volume of worn material is 5 mm3 after 400 m) compared to those obtained with bare titanium alloy substrates (353 Hv0.3, average friction coefficient is 0.57 and a volume of worn material after 400 m is 35 mm3).

Processing of pure Ti by rapid prototyping based on laser cladding

NASA Astrophysics Data System (ADS)

Arias-González, F.; del Val, J.; Comesaña, R.; Lusquiños, F.; Quintero, F.; Riveiro, A.; Boutinguiza, M.; Pou, J.

2013-11-01

Rapid prototyping based on laser cladding is an additive manufacturing (AM) process based on the overlapping of cladding tracks to produce functional components. Powder or wire are fed into a melting pool created using laser radiation as a heat source and the relative movement between the beam and the work piece makes possible to generate pieces layer-by-layer. This technique can be applied for any material which can be melted and the components can be manufactured directly according to a computer aided design (CAD) model. Additive manufacturing is particularly interesting to produce titanium components because, in this case, the loss of material produced by subtractive manufacturing methods is highly costly. Moreover, titanium and its alloys are widely used in biomedical, aircraft, chemical and marine industries due to their biocompatibility, excellent corrosion resistance and superior strength-to-weight ratio. In this research work, a near-infrared laser delivering a maximum power of 500W is used to produce pure titanium thin parts. Dimensions and surface morphology are characterized using Optical Microscopy (OM) and Scanning Electron Microscopy (SEM), the hardness by nanoindentation and the composition by X-Ray Diffraction (XRD) and Energy Dispersive X-Ray Spectroscopy (EDS). The aim of this work is to establish the conditions under which satisfactory properties are obtained and to understand the relationship between microstructure/properties and deposition parameters.

White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

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

Sindelar, R.; Louthan, M.; PNNL, B.

2015-05-29

This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history,more » residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to identical conditions and the material responses to thermo-mechanical exposures will be different depending on the materials and systems used. The discussions at the workshop showed several gaps in the standardization of processes and techniques necessary to assess the long term performance of irradiated zirconium alloy cladding during dry storage and transport. The development of, and adherence to, standards to help bridge these gaps will strengthen the technical basis for long term storage and post-storage operations, provide consistency across the nuclear industry, maximize the value of most observations, and enhance the understanding of behavioral differences among alloys. The need for, and potential benefits of, developing the recommended standards are illustrated in the various sections of this report.« less

Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

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

Ade, Brian J; Bowman, Stephen M; Gauld, Ian C

2015-01-01

[Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, andmore » it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades are inserted in various locations and at varying degrees during BWR operation based on the reload design. The presence of control blades during depletion hardens the neutron spectrum locally due to both moderator displacement and introduction of a thermal neutron absorber. The reactivity impact of control blade presence is investigated herein, as well as the effect of multiple (continuous and intermittent) exposure periods. The coupled effects of control blade presence on power density, void profile, or burnup profile have not been considered to date but will be addressed in future work.« less

Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

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

Ade, Brian J; Marshall, William BJ J; Bowman, Stephen M

2015-01-01

Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technicalmore » basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in various locations and at varying degrees during BWR operation based on the core loading pattern. When present during depletion, control blades harden the neutron spectrum locally because they displace the moderator and absorb thermal neutrons. The investigation of the effect of control blades on post operational cask reactivity is documented herein, as is the effect of multiple (continuous and intermittent) exposure periods with control blades inserted. The coupled effects of control blade presence on power density, void profile, or burnup profile will be addressed in future work.« less

Synthesis and characterization of Ti-27.5Nb alloy made by CLAD® additive manufacturing process for biomedical applications.

PubMed

Fischer, M; Laheurte, P; Acquier, P; Joguet, D; Peltier, L; Petithory, T; Anselme, K; Mille, P

2017-06-01

Biocompatible beta-titanium alloys such as Ti-27.5(at.%)Nb are good candidates for implantology and arthroplasty applications as their particular mechanical properties, including low Young's modulus, could significantly reduce the stress-shielding phenomenon usually occurring after surgery. The CLAD® process is a powder blown additive manufacturing process that allows the manufacture of patient specific (i.e. custom) implants. Thus, the use of Ti-27.5(at.%)Nb alloy formed by CLAD® process for biomedical applications as a mean to increase cytocompatibility and mechanical biocompatibility was investigated in this study. The microstructural properties of the CLAD-deposited alloy were studied with optical microscopy and electron back-scattered diffraction (EBSD) analysis. The conservation of the mechanical properties of the Ti-27.5Nb material after the transformation steps (ingot-powder atomisation-CLAD) were verified with tensile tests and appear to remain close to those of reference material. Cytocompatibility of the material and subsequent cell viability tests showed that no cytotoxic elements are released in the medium and that viable cells proliferated well. Copyright © 2017 Elsevier B.V. All rights reserved.

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

Kohnert, Aaron A.; Dasgupta, Dwaipayan; Wirth, Brian

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensuremore » that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.« less

Cascaded-cladding-pumped cascaded Raman fiber amplifier.

PubMed

Jiang, Huawei; Zhang, Lei; Feng, Yan

2015-06-01

The conversion efficiency of double-clad Raman fiber laser is limited by the cladding-to-core area ratio. To get high conversion efficiency, the inner-cladding-to-core area ratio has to be less than about 8, which limits the brightness enhancement. To overcome the problem, a cascaded-cladding-pumped cascaded Raman fiber laser with multiple-clad fiber as the Raman gain medium is proposed. A theoretical model of Raman fiber amplifier with multiple-clad fiber is developed, and numerical simulation proves that the proposed scheme can improve the conversion efficiency and brightness enhancement of cladding pumped Raman fiber laser.

Discrete parametric band conversion in silicon for mid-infrared applications.

PubMed

Tien, En-Kuang; Huang, Yuewang; Gao, Shiming; Song, Qi; Qian, Feng; Kalyoncu, Salih K; Boyraz, Ozdal

2010-10-11

Silicon photonics has great potential for mid-wave-infrared applications. The dispersion of waveguide can be manipulated by waveguide dimension and cladding materials. Simulation shows that <3 μm wide conversion can be achieved by tuning the pump wavelength.

Advanced materials for space nuclear power systems

NASA Technical Reports Server (NTRS)

Titran, Robert H.; Grobstein, Toni L.; Ellis, David L.

1991-01-01

The overall philosophy of the research was to develop and characterize new high temperature power conversion and radiator materials and to provide spacecraft designers with material selection options and design information. Research on three candidate materials (carbide strengthened niobium alloy PWC-11 for fuel cladding, graphite fiber reinforced copper matrix composites for heat rejection fins, and tungsten fiber reinforced niobium matrix composites for fuel containment and structural supports considered for space power system applications is discussed. Each of these types of materials offers unique advantages for space power applications.

RERTR-12 Post-irradiation Examination Summary Report

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

Rice, Francine; Williams, Walter; Robinson, Adam

2015-02-01

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

Very high numerical aperture light transmitting device

DOEpatents

Allison, Stephen W.; Boatner, Lynn A.; Sales, Brian C.

1998-01-01

A new light-transmitting device using a SCIN glass core and a novel calcium sodium cladding has been developed. The very high index of refraction, radiation hardness, similar solubility for rare earths and similar melt and viscosity characteristics of core and cladding materials makes them attractive for several applications such as high-numerical-aperture optical fibers and specialty lenses. Optical fibers up to 60 m in length have been drawn, and several simple lenses have been designed, ground, and polished. Preliminary results on the ability to directly cast optical components of lead-indium phosphate glass are also discussed as well as the suitability of these glasses as a host medium for rare-earth ion lasers and amplifiers.

Summary of LCRE fuel element design including supporting experimental data

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

None, None

Declassified 18 Sep 1973. The design basis of the LCRE fuel pin is presented. The fuel pin consists of a Cb-1 Zr alloy cladding tube 0.305 inch diameter, 0.015 inch wall thickness and 35.96 inches long. The active fuel section is 13.5 inches long, with top and bottom reflector rods each 6.9 inches long and with a 4 inch gas accumulation space at each end. The cladding is designed as a pressure vessel to contain the gases released from the fuel and end refiector materials, which results in an internal gas pressure buildup in the pins during reactor operation. (23more » referencea) (auth)« less

Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

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

Wilson, C.L.; Hesson, G.M.; Pilger, J.P.

1993-09-01

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuelmore » bundle is cooled.« less

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

Fabrication of locally micro-structured fiber Bragg gratings by fs-laser machining

NASA Astrophysics Data System (ADS)

Dutz, Franz J.; Stephan, Valentin; Marchi, Gabriele; Koch, Alexander W.; Roths, Johannes; Huber, Heinz P.

2018-06-01

Here, we describe a method for producing locally micro-structured fiber Bragg gratings (LMFGB) by fs-laser machining. This technique enables the precise and reproducible ablation of cladding material to create circumferential grooves inside the claddings of optical fibers. From initial ablation experiments we acquired optimized process parameters. The fabricated grooves were located in the middle of uniform type I fiber Bragg gratings. LMFBGs with four different groove widths of 48, 85, 135 and 205 μ { {m}} were produced. The grooves exhibited constant depths of about 30 μ {m} and steep sidewall angles. With the combination of micro-structures and fiber Bragg gratings, fiber optic sensor elements with enhanced functionalities can be achieved.

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

Clarke, Kester Diederik

The intent of this report is to document a procedure used at LANL for HIP bonding aluminum cladding to U-10Mo fuel foils using a formed HIP can for the Domestic Reactor Conversion program in the NNSA Office of Material, Management and Minimization, and provide some details that may not have been published elsewhere. The HIP process is based on the procedures that have been used to develop the formed HIP can process, including the baseline process developed at Idaho National Laboratory (INL). The HIP bonding cladding process development is summarized in the listed references. Further iterations with Babcock & Wilcoxmore » (B&W) to refine the process to meet production and facility requirements is expected.« less

Coatings Extend Life of Engines and Infrastructure

NASA Technical Reports Server (NTRS)

2010-01-01

MesoCoat Inc., of Euclid, Ohio, collaborated with Glenn Research Center to provide thermal barrier coating (TBC) technology, developed by Glenn researcher Dongming Zhu, to enhance the lifespan and performance of engines in U.S. Air Force legacy aircraft. The TBC reduces thermal stresses on engine parts, increasing component life by 50 percent. MesoCoat is also producing metal cladding technology that may soon provide similar life-lengthening benefits for the Nation's infrastructure. Through a Space Act Agreement with Glenn, the company employs the Center's high-density infrared arc lamp system to bond its cladding materials for demonstration prototypes; the coating technology can prevent corrosion on metal beams, pipes, and rebar for up to 100 years.

Process for producing clad superconductive materials

DOEpatents

Cass, Richard B.; Ott, Kevin C.; Peterson, Dean E.

1992-01-01

A process for fabricating superconducting composite wire by the steps of placing a superconductive precursor admixture capable of undergoing a self propagating combustion in stoichiometric amounts sufficient to form a superconductive product within a metal tube, sealing one end of said tube, igniting said superconductive precursor admixture whereby said superconductive precursor admixture endburns along the length of the admixture, and cross-section reducing said tube at a rate substantially equal to the rate of burning of said superconductive precursor admixture and at a point substantially planar with the burnfront of the superconductive precursor mixture, whereby a clad superconductive product is formed in situ, the product characterized as superconductive without a subsequent sintering stage, is disclosed.

Critical cladding radius for hybrid cladding modes

NASA Astrophysics Data System (ADS)

Guyard, Romain; Leduc, Dominique; Lupi, Cyril; Lecieux, Yann

2018-05-01

In this article we explore some properties of the cladding modes guided by a step-index optical fiber. We show that the hybrid modes can be grouped by pairs and that it exists a critical cladding radius for which the modes of a pair share the same electromagnetic structure. We propose a robust method to determine the critical cladding radius and use it to perform a statistical study on the influence of the characteristics of the fiber on the critical cladding radius. Finally we show the importance of the critical cladding radius with respect to the coupling coefficient between the core mode and the cladding modes inside a long period grating.

Effects of materials and design on the criticality and shielding assessment of canister concepts for the disposal of spent nuclear fuel.

PubMed

Gutiérrez, Miguel Morales; Caruso, Stefano; Diomidis, Nikitas

2018-05-19

According to the Swiss disposal concept, the safety of a deep geological repository for spent nuclear fuel (SNF) is based on a multi-barrier system. The disposal canister is an important component of the engineered barrier system, aiming to provide containment of the SNF for thousands of years. This study evaluates the criticality safety and shielding of candidate disposal canister concepts, focusing on the fulfilment of the sub-criticality criterion and on limiting radiolysis processes at the outer surface of the canister which can enhance corrosion mechanisms. The effective neutron multiplication factor (k-eff) and the surface dose rates are calculated for three different canister designs and material combinations for boiling water reactor (BWR) canisters, containing 12 spent fuel assemblies (SFA), and pressurized water reactor (PWR) canisters, with 4 SFAs. For each configuration, individual criticality and shielding calculations were carried out. The results show that k-eff falls below the defined upper safety limit (USL) of 0.95 for all BWR configurations, while staying above USL for the PWR ones. Therefore, the application of a burnup credit methodology for the PWR case is required, being currently under development. Relevant is also the influence of canister material and internal geometry on criticality, enabling the identification of safer fuel arrangements. For a final burnup of 55MWd/kgHM and 30y cooling time, the combined photon-neutron surface dose rate is well below the threshold of 1 Gy/h defined to limit radiation-induced corrosion of the canister in all cases. Copyright © 2018 Elsevier Ltd. All rights reserved.

Lung volumes predict survival in patients with chronic lung allograft dysfunction.

PubMed

Kneidinger, Nikolaus; Milger, Katrin; Janitza, Silke; Ceelen, Felix; Leuschner, Gabriela; Dinkel, Julien; Königshoff, Melanie; Weig, Thomas; Schramm, René; Winter, Hauke; Behr, Jürgen; Neurohr, Claus

2017-04-01

Identification of disease phenotypes might improve the understanding of patients with chronic lung allograft dysfunction (CLAD). The aim of the study was to assess the impact of pulmonary restriction and air trapping by lung volume measurements at the onset of CLAD.A total of 396 bilateral lung transplant recipients were analysed. At onset, CLAD was further categorised based on plethysmography. A restrictive CLAD (R-CLAD) was defined as a loss of total lung capacity from baseline. CLAD with air trapping (AT-CLAD) was defined as an increased ratio of residual volume to total lung capacity. Outcome was survival after CLAD onset. Patients with insufficient clinical information were excluded (n=95).Of 301 lung transplant recipients, 94 (31.2%) developed CLAD. Patients with R-CLAD (n=20) and AT-CLAD (n=21), respectively, had a significantly worse survival (p<0.001) than patients with non-R/AT-CLAD. Both R-CLAD and AT-CLAD were associated with increased mortality when controlling for multiple confounding variables (hazard ratio (HR) 3.57, 95% CI 1.39-9.18; p=0.008; and HR 2.65, 95% CI 1.05-6.68; p=0.039). Furthermore, measurement of lung volumes was useful to identify patients with combined phenotypes.Measurement of lung volumes in the long-term follow-up of lung transplant recipients allows the identification of patients who are at risk for worse outcome and warrant special consideration. Copyright ©ERS 2017.

BWRVIP-140NP: BWR Vessel and Internals Project Fracture Toughness and Crack Growth Program on Irradiated Austenitic Stainless Steel

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

Gilman, J

2005-03-15

To prepare for this project, EPRI and BWRVIP conducted a workshop at Ponte Vedra Beach, Florida during February 19-21, 2003 (EPRI report 1007822). Attendees were invited to exchange relevant information on the effects of irradiation on austenitic materials in light water reactors and to produce recommendations for further work. EPRI reviewed the data, recommendations, and conclusions derived from the workshop and developed prioritized test matrices defining new data needs. Proposals were solicited, and selected proposals are the basis for the program described in this report. Results The planned test matrix for fracture toughness testing includes 21 tests on 5 materials.

Composite ceramic superconducting wires for electric motor applications

NASA Astrophysics Data System (ADS)

Halloran, John W.

1988-12-01

This is the Second Quarterly report on a project to develop HTSC wire for an HTSC motor. The raw material for fiber production is an improved YBa2Cu3O(7-x) powder. Continuous spools of green YBa2Cu3O(7-x) fiber are being produced. The major effort in fiber spinning is aimed at improving fiber quality and reducing fiber. Binder burnout and sintering has been intensively investigated. Fiber sintering fibers is done by the rapid zone sintering method. A continuous furnace received near the end of this Quarter will be used for continuous sintering. Continuous silver coated green fiber are produced. We have made progress toward continuous cladding using the mechanical cladding concept. The melt spinning process was successfully applied to YBa2Cu3O(7-x) powders at 50 vol percent solids loadings. The cladding work centered on mechanical cladding of silver treated filaments by solder bonding to copper strips. Aluminum deposits on YBa2Cu3O(7-x) filament surfaces were produced by MOCVD at ATM, but the superconductivity was degraded. Electrical characterization work focused on methods of making low resistance contacts on YBa2Cu3O(7-x) filaments. Emerson Motor Division has begun work on DC heteropolar and homopolar motor designs. The mechanical stresses on conventional copper wires during winding have been characterized to determine the mechanical parameters of motor building.

Initial and Long-Term Movement of Cladding Installed Over Exterior Rigid Insulation

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

Baker, P.

Changes in the International Energy Conservation Code (IECC) from 2009 to 2012 have resulted in the use of exterior rigid insulation becoming part of the prescriptive code requirements. With more jurisdictions adopting the 2012 IECC builders are going to finding themselves required to incorporate exterior insulation in the construction of their exterior wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location. However, there has beenmore » a significant resistance to its widespread implementation due to a lack of research and understanding of the mechanisms involved and potential creep effects of the assembly under the sustained dead load of a cladding. This research was an extension on previous research conducted by BSC in 2011, and 2012. Each year the understanding of the system discrete load component interactions, as well as impacts of environmental loading has increased. The focus of the research was to examine more closely the impacts of screw fastener bending on the total system capacity, effects of thermal expansion and contraction of materials on the compressive forces in the assembly, as well as to analyze a full years worth of cladding movement data from assemblies constructed in an exposed outdoor environment.« less

Laser cladding of tungsten carbides (Spherotene ®) hardfacing alloys for the mining and mineral industry

NASA Astrophysics Data System (ADS)

Amado, J. M.; Tobar, M. J.; Alvarez, J. C.; Lamas, J.; Yáñez, A.

2009-03-01

The abrasive nature of the mechanical processes involved in mining and mineral industry often causes significant wear to the associated equipment and derives non-negligible economic costs. One of the possible strategies to improve the wear resistance of the various components is the deposition of hardfacing layers on the bulk parts. The use of high power lasers for hardfacing (laser cladding) has attracted a great attention in the last decade as an alternative to other more standard methods (arc welding, oxy-fuel gas welding, thermal spraying). In laser cladding the hardfacing material is used in powder form. For high hardness applications Ni-, Co- or Fe-based alloys containing hard phase carbides at different ratios are commonly used. Tungsten carbides (WC) can provide coating hardness well above 1000 HV (Vickers). In this respect, commercially available WC powders normally contain spherical micro-particles consisting of crushed WC agglomerates. Some years ago, Spherotene ® powders consisting of spherical-fused monocrystaline WC particles, being extremely hard, between 1800 and 3000 HV, were patented. Very recently, mixtures of Ni-based alloy with Spherotene powders optimized for laser processing were presented (Technolase ®). These mixtures have been used in our study. Laser cladding tests with these powders were performed on low carbon steel (C25) substrates, and results in terms of microstructure and hardness will be discussed.

Conservation of Stone Cladding on the FAÇADE of Royal Palace in Caserta

NASA Astrophysics Data System (ADS)

Titomanlio, I.

2013-07-01

The beauty of cultural heritage and monumental architecture, is often linked to their non-structural elements and decorative stones façades cladding. The collapse of these elements causes significant consequences that interest the social, the economic, the historical and the technical fields. Several regulatory documents and literature studies contain methods to address the question of relief and of the risk analysis and due to the non - structural stones security. Among the references are widespread international regulatory documents prepared by the Federal Emergency Management Agency of the United States by Applied Technology Council and California. In Italy there are some indications contained in the Norme Tecniche per le Costruzioni and the Direttiva del Presidente del Consiglio dei Ministri in 2007, finalize to the reduction of seismic risk assessment of cultural heritage. The paper, using normative references and scientific researches, allows to analyze on Royal Palace of Caserta the safety and the preservation of cultural heritage and the vulnerability of non-structural stones façade cladding. Using sophisticated equipments of Laboratory ARS of the Second University of Naples, it was possible to analyze the collapse of stone elements due to degradation caused by natural phenomena of deterioration (age of the building, type of materials, geometries , mode of fixing of the elements themselves). The paper explains the collapse mechanisms of stones façade cladding of Luigi Vanvitelli Palace.

Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl Alloys

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

Field, Kevin G; Yamamoto, Yukinori; Pint, Bruce A

2016-01-01

A large effort is underway under the leadership of US DOE Fuel Cycle R&D program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanicalmore » characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs.« less

JPRS Report, Science & Technology, USSR: Materials Science

DTIC Science & Technology

1988-01-08

ceramic product was found to have a high ratio of thickness-to-radial electromechanical coiipling coefficients. Dielectric permittivity and loss...equipment lacked just metal-cladding lubricants! The colossal capacities of metal-cutting equipment and tools, i.e. the foundation of all machine building

All fiber cladding mode stripper with uniform heat distribution and high cladding light loss manufactured by CO2 laser ablation

NASA Astrophysics Data System (ADS)

Jebali, M. A.; Basso, E. T.

2018-02-01

Cladding mode strippers are primarily used at the end of a fiber laser cavity to remove high-power excess cladding light without inducing core loss and beam quality degradation. Conventional manufacturing methods of cladding mode strippers include acid etching, abrasive blasting or laser ablation. Manufacturing of cladding mode strippers using laser ablation consist of removing parts of the cladding by fused silica ablation with a controlled penetration and shape. We present and characterize an optimized cladding mode stripper design that increases the cladding light loss with a minimal device length and manufacturing time. This design reduces the localized heat generation by improving the heat distribution along the device. We demonstrate a cladding mode stripper written on a 400um fiber with cladding light loss of 20dB, with less than 0.02dB loss in the core and minimal heating of the fiber and coating. The manufacturing process of the designed component is fully automated and takes less than 3 minutes with a very high throughput yield.

Femtosecond laser inscription of optical circuits in the cladding of optical fibers

NASA Astrophysics Data System (ADS)

Grenier, Jason R.

The aim of this dissertation was to address the question of whether the cladding of single-mode fibers (SMFs) could be modified to enable optical fibers to serve as a more integrated, highly functional platform for optical circuit devices that can efficiently interconnect with the pre-existing fiber core waveguide. The approach adopted in this dissertation was to employ femtosecond laser direct writing (FLDW), an inherently 3D fabrication technique that harnesses non-linear laser-material interactions to modify the fused silica fiber cladding. A fiber mounting and alignment technique was developed along with oil-immersion focusing to address the strong aberrations caused by the cylindrical fiber shape. The development of real-time device monitoring during the FLDW was instrumental to overcome the acute coupling sensitivity to laser alignment errors of +/-1 ?m positional uncertainty, and thereby opened a new practical direction for the precise fabrication of optical devices inside optical fibers. These powerful and flexible laser fabrication and characterization techniques were successfully employed to optimize optical waveguiding devices positioned within the core and cladding of optical fibers. X-, S-Bend, and directional couplers were developed to enable efficient coupling between the laser-formed cladding devices and the pre-existing core waveguide, enabling up to 62% power transfer over bandwidths up to 300 nm at telecommunication wavelengths. Precise alignment of femtosecond laser modification tracks were positioned inside or near the core waveguide of SMFs was further shown to enable a flexible reshaping of the optical properties to create multimode guiding sections arbitrarily along the fiber length. This core waveguide modification facilitated the precise formation of multimode interferometers along the core waveguide to precisely tailor the modal profiles, and control the spectral and polarization response. In-fiber multimode interference (MMI) splitters and couplers were fabricated with coupling ratios from 2% to 50% over a broad 350 nm bandwidth across the telecommunication band. Laser-induced birefringence was harnessed to generate polarization dependent MMI devices for strong polarization filtering (24 dB isolation), or polarization selective taps with up to 50% tapping efficiency over a 25 nm bandwidth. This dissertation is therefore the first demonstration of femtosecond laser direct writing as a flexible and monolithic means of embedding and integrating highly functional optical circuit devices within the cladding of optical fibers that can interconnect efficiently with the pre-existing fiber core waveguide. These developments represent a significant technological advancement for creating new 3D photonic integrated microsystems within the cladding of optical fibers and underpins a new technological platform of fiber cladding photonics.

Microstructure and Mechanical Properties of Laser Clad and Post-cladding Tempered AISI H13 Tool Steel

NASA Astrophysics Data System (ADS)

Telasang, Gururaj; Dutta Majumdar, Jyotsna; Wasekar, Nitin; Padmanabham, G.; Manna, Indranil

2015-05-01

This study reports a detailed investigation of the microstructure and mechanical properties (wear resistance and tensile strength) of hardened and tempered AISI H13 tool steel substrate following laser cladding with AISI H13 tool steel powder in as-clad and after post-cladding conventional bulk isothermal tempering [at 823 K (550 °C) for 2 hours] heat treatment. Laser cladding was carried out on AISI H13 tool steel substrate using a 6 kW continuous wave diode laser coupled with fiber delivering an energy density of 133 J/mm2 and equipped with a co-axial powder feeding nozzle capable of feeding powder at the rate of 13.3 × 10-3 g/mm2. Laser clad zone comprises martensite, retained austenite, and carbides, and measures an average hardness of 600 to 650 VHN. Subsequent isothermal tempering converted the microstructure into one with tempered martensite and uniform dispersion of carbides with a hardness of 550 to 650 VHN. Interestingly, laser cladding introduced residual compressive stress of 670 ± 15 MPa, which reduces to 580 ± 20 MPa following isothermal tempering. Micro-tensile testing with specimens machined from the clad zone across or transverse to cladding direction showed high strength but failure in brittle mode. On the other hand, similar testing with samples sectioned from the clad zone parallel or longitudinal to the direction of laser cladding prior to and after post-cladding tempering recorded lower strength but ductile failure with 4.7 and 8 pct elongation, respectively. Wear resistance of the laser surface clad and post-cladding tempered samples (evaluated by fretting wear testing) registered superior performance as compared to that of conventional hardened and tempered AISI H13 tool steel.

Realization of optical multimode TSV waveguides for Si-Interposer in 3D-chip-stacks

NASA Astrophysics Data System (ADS)

Killge, S.; Charania, S.; Richter, K.; Neumann, N.; Al-Husseini, Z.; Plettemeier, D.; Bartha, J. W.

2017-05-01

Optical connectivity has the potential to outperform copper-based TSVs in terms of bandwidth at the cost of more complexity due to the required electro-optical and opto-electrical conversion. The continuously increasing demand for higher bandwidth pushes the breakeven point for a profitable operation to shorter distances. To integrate an optical communication network in a 3D-chip-stack optical through-silicon vertical VIAs (TSV) are required. While the necessary effort for the electrical/optical and vice versa conversion makes it hard to envision an on-chip optical interconnect, a chip-to-chip optical link appears practicable. In general, the interposer offers the potential advantage to realize electro-optical transceivers on affordable expense by specific, but not necessarily CMOS technology. We investigated the realization and characterization of optical interconnects as a polymer based waveguide in high aspect ratio (HAR) TSVs proved on waferlevel. To guide the optical field inside a TSV as optical-waveguide or fiber, its core has to have a higher refractive index than the surrounding material. Comparing different material / technology options it turned out that thermal grown silicon dioxide (SiO2) is a perfect candidate for the cladding (nSiO2 = 1.4525 at 850 nm). In combination with SiO2 as the adjacent polymer layer, the negative resist SU-8 is very well suited as waveguide material (nSU-8 = 1.56) for the core. Here, we present the fabrication of an optical polymer based multimode waveguide in TSVs proved on waferlevel using SU-8 as core and SiO2 as cladding. The process resulted in a defect-free filling of waveguide TSVs with SU-8 core and SiO2 cladding up to aspect ratio (AR) 20:1 and losses less than 3 dB.

The advances and characteristics of high-power diode laser materials processing

NASA Astrophysics Data System (ADS)

Li, Lin

2000-10-01

This paper presents a review of the direct applications of high-power diode lasers for materials processing including soldering, surface modification (hardening, cladding, glazing and wetting modifications), welding, scribing, sheet metal bending, marking, engraving, paint stripping, powder sintering, synthesis, brazing and machining. The specific advantages and disadvantages of diode laser materials processing are compared with CO 2, Nd:YAG and excimer lasers. An effort is made to identify the fundamental differences in their beam/material interaction characteristics and materials behaviour. Also an appraisal of the future prospects of the high-power diode lasers for materials processing is given.

A high converter concept for fuel management with blanket fuel assemblies in boiling water reactors

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

Martinez-Frances, N.; Timm, W.; Rossbach, D.

2012-07-01

Studies on the natural Uranium saving and waste reduction potential of a multiple-plant BWR system were performed. The BWR High Converter system should enable a multiple recycling of MOX fuel in current BWR plants by introducing blanket fuel assemblies and burning Uranium and MOX fuel separately. The feasibility of Uranium cores with blankets and full-MOX cores with Plutonium qualities as low as 40% were studied. The power concentration due to blanket insertion is manageable with modern fuel and acceptable values for the thermal limits and reactivity coefficients were obtained. While challenges remain, full-MOX cores also complied with the main designmore » criteria. The combination of Uranium and Plutonium burners in appropriate proportions could enable obtaining as much as 40% more energy out of Uranium ore. Moreover, a proper adjustment of blanket average stay and Plutonium qualities could lead to a system with nearly no Plutonium left for final disposal. The achievement of such goals with current light water technology makes the BWR HC concept an attractive option to improve the fuel cycle until Gen-IV designs are mature. (authors)« less

Investigation of two and three parameter equations of state for cryogenic fluids

NASA Technical Reports Server (NTRS)

Jenkins, Susan L.; Majumdar, Alok K.; Hendricks, Robert C.

1990-01-01

Two-phase flows are a common occurrence in cryogenic engines and an accurate evaluation of the heat-transfer coefficient in two-phase flow is of significant importance in their analysis and design. The thermodynamic equation of state plays a key role in calculating the heat transfer coefficient which is a function of thermodynamic and thermophysical properties. An investigation has been performed to study the performance of two- and three-parameter equations of state to calculate the compressibility factor of cryogenic fluids along the saturation loci. The two-parameter equations considered here are van der Waals and Redlich-Kwong equations of state. The three-parameter equation represented here is the generalized Benedict-Webb-Rubin (BWR) equation of Lee and Kesler. Results have been compared with the modified BWR equation of Bender and the extended BWR equations of Stewart. Seven cryogenic fluids have been tested; oxygen, hydrogen, helium, nitrogen, argon, neon, and air. The performance of the generalized BWR equation is poor for hydrogen and helium. The van der Waals equation is found to be inaccurate for air near the critical point. For helium, all three equations of state become inaccurate near the critical point.

High temperature gradient cobalt based clad developed using microwave hybrid heating

NASA Astrophysics Data System (ADS)

Prasad, C. Durga; Joladarashi, Sharnappa; Ramesh, M. R.; Sarkar, Anunoy

2018-04-01

The development of cobalt based cladding on a titanium substrate using microwave cladding technique is benchmark in coating area. The developed cladding would serve the function of a corrosion resistant coating under high temperatures. Clads of thickness 500 µm have been developed by microwave hybrid heating. A microwave furnace of 2.45GHz frequency was used at a 900W power level for processing. Impact of processing time on melting and adhesion of clad has been discussed. The study also extended to static thermal analysis of simple parts with cladding using commercial Finite Element analysis (FEA) software. A comparative study is explored between four variants of the clad being developed. The analysis has been conducted using a square sample. Similar temperature gradient is also shown for a proposed multi-layer coating, which includes a thermal barrier coating yttria stabilized zirconia (YSZ) on top of the corrosion resistant clad. The YSZ coating would protect the corrosion resistant cladding and substrate from high temperatures.

Effect of Mo on Microstructures and Wear Properties of In Situ Synthesized Ti(C,N)/Ni-Based Composite Coatings by Laser Cladding.

PubMed

Wu, Fan; Chen, Tao; Wang, Haojun; Liu, Defu

2017-09-06

Using Ni60 alloy, C, TiN and Mo mixed powders as the precursor materials, in situ synthesized Ti(C,N) particles reinforcing Ni-based composite coatings are produced on Ti6Al4V alloys by laser cladding. Phase constituents, microstructures and wear properties of the composite coatings with 0 wt % Mo, 4 wt % Mo and 8 wt % Mo additions are studied comparatively. Results indicate that Ti(C,N) is formed by the in situ metallurgical reaction, the (Ti,Mo)(C,N) rim phase surrounding the Ti(C,N) ceramic particle is synthesized with the addition of Mo, and the increase of Mo content is beneficial to improve the wear properties of the cladding coatings. Because of the effect of Mo, the grains are remarkably refined and a unique core-rim structure that is uniformly dispersed in the matrix appears; meanwhile, the composite coatings with Mo addition exhibit high hardness and excellent wear resistance due to the comprehensive action of dispersion strengthening, fine grain strengthening and solid solution strengthening.

Effect of Mo on Microstructures and Wear Properties of In Situ Synthesized Ti(C,N)/Ni-Based Composite Coatings by Laser Cladding

PubMed Central

Chen, Tao; Wang, Haojun

2017-01-01

Using Ni60 alloy, C, TiN and Mo mixed powders as the precursor materials, in situ synthesized Ti(C,N) particles reinforcing Ni-based composite coatings are produced on Ti6Al4V alloys by laser cladding. Phase constituents, microstructures and wear properties of the composite coatings with 0 wt % Mo, 4 wt % Mo and 8 wt % Mo additions are studied comparatively. Results indicate that Ti(C,N) is formed by the in situ metallurgical reaction, the (Ti,Mo)(C,N) rim phase surrounding the Ti(C,N) ceramic particle is synthesized with the addition of Mo, and the increase of Mo content is beneficial to improve the wear properties of the cladding coatings. Because of the effect of Mo, the grains are remarkably refined and a unique core-rim structure that is uniformly dispersed in the matrix appears; meanwhile, the composite coatings with Mo addition exhibit high hardness and excellent wear resistance due to the comprehensive action of dispersion strengthening, fine grain strengthening and solid solution strengthening. PMID:28878190

Temperature and composition profile during double-track laser cladding of H13 tool steel

NASA Astrophysics Data System (ADS)

He, X.; Yu, G.; Mazumder, J.

2010-01-01

Multi-track laser cladding is now applied commercially in a range of industries such as automotive, mining and aerospace due to its diversified potential for material processing. The knowledge of temperature, velocity and composition distribution history is essential for a better understanding of the process and subsequent microstructure evolution and properties. Numerical simulation not only helps to understand the complex physical phenomena and underlying principles involved in this process, but it can also be used in the process prediction and system control. The double-track coaxial laser cladding with H13 tool steel powder injection is simulated using a comprehensive three-dimensional model, based on the mass, momentum, energy conservation and solute transport equation. Some important physical phenomena, such as heat transfer, phase changes, mass addition and fluid flow, are taken into account in the calculation. The physical properties for a mixture of solid and liquid phase are defined by treating it as a continuum media. The velocity of the laser beam during the transition between two tracks is considered. The evolution of temperature and composition of different monitoring locations is simulated.

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

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

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

Ade, Brian J; Marshall, William BJ J; Martinez-Gonzalez, Jesus S

Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents themore » analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.« less

Steam Oxidation Testing in the Severe Accident Test Station

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

Pint, Bruce A.

After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative wouldmore » significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.« less

HRTEM and chemical study of an ion-irradiated chromium/zircaloy-4 interface

NASA Astrophysics Data System (ADS)

Wu, A.; Ribis, J.; Brachet, J.-C.; Clouet, E.; Leprêtre, F.; Bordas, E.; Arnal, B.

2018-06-01

Chromium-coated zirconium alloys are being studied as Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors (LWRs). Those materials are especially studied to improve the oxidation resistance of LWRs current fuel claddings in nominal and at High Temperature (HT) for hypothetical accidental conditions such as LOss of Coolant Accident. Beyond their HT behavior, it is essential to assess the materials behavior under irradiation. A first generation chromium/Zircaloy-4 interface was thus irradiated with 20 MeV Kr8+ ions at 400 °C up to 10 dpa. High-Resolution Transmission Electron Microscopy and chemical analysis (EDS) were conducted at the Cr/Zr interface. The atomic structure of the interface reveals the presence of Zr(Fe, Cr)2 Laves phase, displaying both C14 and C15 structure. After irradiation, only the C14 structure was observed and atomic row matching was preserved across the different interfaces, thus ensuring a good adhesion of the coating after irradiation.

Microstructural Evolution of Inconel 625 and Inconel 686CPT Weld Metal for Clad Carbon Steel Linepipe Joints: A Comparator Study

NASA Astrophysics Data System (ADS)

Maltin, Charles A.; Galloway, Alexander M.; Mweemba, Martin

2014-07-01

Microstructural evolution of Inconel 625 and Inconel 686CPT filler metals, used for the fusion welding of clad carbon steel linepipe, has been investigated and compared. The effects of iron dilution from the linepipe parent material on the elemental segregation potential of the filler metal chemistry have been considered. The results obtained provide significant evidence to support the view that, in Inconel 686CPT weld metal, the segregation of tungsten is a function of the level of iron dilution from the parent material. The data presented indicate that the incoherent phase precipitated in the Inconel 686CPT weld metal has a morphology that is dependent on tungsten enrichment and, therefore, iron dilution. Furthermore, in the same weld metal, a continuous network of finer precipitates was observed. The Charpy impact toughness of each filler metal was evaluated, and the results highlighted the superior impact toughness of the Inconel 625 weld metal over that of Inconel 686CPT.

Studies of thermionic materials for space power applications

NASA Technical Reports Server (NTRS)

1972-01-01

The effect of microstructures of tungsten cladding on the transport rates of carbide fuel components was studied at 2073 K. hyperstoichiometric 90UC-10ZrC containing 4 wt% tungsten was clad with six types of tungsten material of 40 mil thickness. Screening tests of 1000 hours were carried out, and then selected samples were subjected to long-term tests up to 10,000 hours. The results indicate that the microstructures strongly affect the transport rates of carbide fuel components. The conditions for preparing (110) oriented cylindrical chloride tungsten emitters of high vacuum work functions were also investigated. Specimen sets were deposited on fluoride tungsten substrates for evaluating the effects of various deposition parameters on the degree and uniformity of the (110) preferred orientation and the vacuum work function. Long-term tests showed that the high vacuum work function of a cylindrical emitter was stable and the chloride tungsten to fluoride tungsten bond remained in excellent shape after 4850 hours at 2073 K.

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

Maloy, Stuart Andrew; Pestovich, Kimberly Shay; Anderoglu, Osman

The Fuel Cycle Research and Development program is investigating methods of transmuting minor actinides in various fuel cycle options. To achieve this goal, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Recent results from testing numerous ferritic/martensitic steels at low temperatures suggest that improvements inmore » low temperature radiation tolerance can be achieved through carefully controlling the nitrogen content in these alloys. Thus, four new heats of HT-9 were produced with controlled nitrogen content: two by Metalwerks and two by Sophisticated Alloys. Initial results on these new alloys are presented including microstructural analysis and hardness testing. Future testing will include irradiation testing with ions and in reactor.« less

Microstructured optical fibers for gas sensing systems

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

Challener, William Albert; Choudhury, Niloy; Palit, Sabarni

2017-10-17

Microstructured optical fiber (MOF) includes a cladding extending a length between first and second ends. The cladding includes an inner porous microstructure that at least partially surrounds a hollow core. A perimeter contour of the hollow core has a non-uniform radial distance from a center axis of the cladding such that first segments of the cladding along the perimeter contour have a shorter radial distance from the center axis relative to second segments of the cladding along the perimeter contour. The cladding receives and propagates light energy through the hollow core, and the inner porous microstructure substantially confines the lightmore » energy within the hollow core. The cladding defines at least one port hole that extends radially from an exterior surface of the cladding to the hollow core. Each port hole penetrates the perimeter contour of the hollow core through one of the second segments of the cladding.« less

FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Shaner, B.E.

1961-08-15

The fissile material consists of about 64 to 70% (weight) zirconium dioxide, 15 to 19% uranium dioxide, and 8 to 17% calcium oxide. The fissile material is formed into sintered composites which are disposed in a compartmented fuel element, comprising essentially a flat filler plate having a plurality of compartments therein, enclosed in cladding plates of the same material as the filler plate. The resultant fuel has good resistance to corrosion in high temperature pressurized water, good dimensional stability to elevated temperatures, and good resistance to thermal shock. (AEC)

Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

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

Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph

2015-09-01

It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T 2O. In a standard processing flowsheet, tritium management would bemore » accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. Samples of Surry-2 and H.B. Robinson pressurized water reactor cladding were heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. Cladding samples were also heated within the temperature range of 480–600ºC expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. The tritium content of the Surry-2 and H.B. Robinson cladding was measured to be ~234 and ~500 µCi/g, respectively. Heating the Surry-2 cladding at 500°C for 24 h removed ~0.2% of the tritium from the cladding, and heating at 700°C for 24 h removed ~9%. Heating the H.B. Robinson cladding at 700°C for 24 h removed ~11% of the tritium. When samples of the Surry-2 and H.B. Robinson claddings were heated at 700°C for 96 h, essentially all of the tritium in the cladding was removed. However, only ~3% of the tritium was removed when a sample of Surry-2 cladding was heated at 600°C for 96 h. These data indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in operations involving cladding recycle.« less

Influence of adding strong-carbide-formation elements multiply on particle-reinforced Fe-matrix composite layer produced by laser cladding

NASA Astrophysics Data System (ADS)

Ma, Mingxing; Liu, Wenjin; Zhong, Minlin; Zhang, Hongjun; Zhang, Weiming

2005-01-01

In the research hotspot of particle reinforced metal-matrix composite layer produced by laser cladding, in-situ reinforced particles obtained by adding strong-carbide-formation elements into cladding power have been attracting more attention for their unique advantage. The research has demonstrated that when adding strong-carbide-formation elements-Ti into the cladding powder of the Fe-C-Si-B separately, by optimizing the composition, better cladding coating with the characters of better strength and toughness, higher wear resistance and free of cracks. When the microstructure of cladding coating is hypoeutectic microstructure, its comprehensive performance is best. The research discovered that, compositely adding the strong-carbide-formation elements like Ti+V, Ti+Zr or V+Zr into the cladding coating is able to improve its comprehensive capability. All the cladding coatings obtained are hypoeutectic microstructure. The cladding coatings have a great deal of particulates, and its average microhardness reaches HV0.2700-1400. The research also discovered that the cladding coating obtained is of less cracking after adding the Ti+Zr.

Energetic additive manufacturing process with feed wire

DOEpatents

Harwell, Lane D.; Griffith, Michelle L.; Greene, Donald L.; Pressly, Gary A.

2000-11-07

A process for additive manufacture by energetic wire deposition is described. A source wire is fed into a energy beam generated melt-pool on a growth surface as the melt-pool moves over the growth surface. This process enables the rapid prototyping and manufacture of fully dense, near-net shape components, as well as cladding and welding processes. Alloys, graded materials, and other inhomogeneous materials can be grown using this process.

Organic-inorganic hybrid material SUNCONNECT® for photonic integrated circuit

NASA Astrophysics Data System (ADS)

Nawata, Hideyuki; Oshima, Juro; Kashino, Tsubasa

2018-02-01

In this paper, we report the feature and properties about organic-inorganic hybrid material, "SUNCONNECT®" for photonic integrated circuit. "SUNCONNECT®" materials have low propagation loss at 1310nm (0.29dB/cm) and 1550nm (0.45dB/cm) respectively. In addition, the material has high thermal resistance both high temperature annealing test at 300°C and also 260°C solder heat resistance test. For actual device application, high reliability is required. 85°C /85% test was examined by using multi-mode waveguide. As a result, it indicated that variation of insertion loss property was not changed significantly after high temperature / high humidity test. For the application to photonic integrated circuit, it was demonstrated to fabricate polymer optical waveguide by using three different methods. Single-micron core pattern can be fabricated on cladding layer by using UV lithography with proximity gap exposure. Also, single-mode waveguide can be also fabricated with over cladding. On the other hands, "Mosquito method" and imprint method can be applied to fabricate polymer optical waveguide. Remarkably, these two methods can fabricate gradedindex type optical waveguide without using photo mask. In order to evaluate the optical performance, NFP's observation, measurement of insertion loss and propagation loss by cut-back methods were carried out by using each waveguide sample.

Coupled field effects in BWR stability simulations using SIMULATE-3K

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

Borkowski, J.; Smith, K.; Hagrman, D.

1996-12-31

The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17.

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

NASA Astrophysics Data System (ADS)

Mella, R.; Wenman, M. R.

2013-06-01

Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of different FE and finite difference models. Non-linear mechanical behaviour of the fuel and cladding including, fuel creep and swelling and cladding creep and plasticity each with dependencies on a variety of different properties. A fission gas release model which takes inputs from first principles calculations. Explicitly integrated inventory calculations performed in a coupled manner. Freedom to model steady state and transient behaviour using implicit time integration. The whole pin geometry is considered over an entire typical fuel life. The model showed by examination of normal operation and a subsequent transient chosen for software demonstration purposes: ABAQUS may be a sufficiently flexible platform to develop a complete and verified fuel performance code. The importance and effectiveness of the geometry of the fuel spacer pellets was characterised. The fuels performance under normal conditions (high friction no power spikes) would not suggest serious degradation of the cladding in fuel life. Large plastic strains were found when pellet bonding was strong, these would appear at all pellets cladding triple points and all pellet radial crack and cladding interfaces thus showing a possible axial direction to cracks forming from ductility exhaustion.

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

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

Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.« less

Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions

NASA Astrophysics Data System (ADS)

Ott, L. J.; Robb, K. R.; Wang, D.

2014-05-01

Following the severe accidents at the Japanese Fukushima Daiichi Nuclear Power Station in 2011, the US Department of Energy initiated research and development on the enhancement of the accident tolerance of light water reactors by the development of fuels/cladding that, in comparison with the standard UO2/Zircaloy (Zr) system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. Analyses are presented that illustrate the impact of these new candidate fuel/cladding materials on the fuel performance at normal operating conditions and on the reactor system under DB and BDB accident conditions.

Solar Power Wires Based on Organic Photovoltaic Materials

NASA Astrophysics Data System (ADS)

Lee, Michael R.; Eckert, Robert D.; Forberich, Karen; Dennler, Gilles; Brabec, Christoph J.; Gaudiana, Russell A.

2009-04-01

Organic photovoltaics in a flexible wire format has potential advantages that are described in this paper. A wire format requires long-distance transport of current that can be achieved only with conventional metals, thus eliminating the use of transparent oxide semiconductors. A phase-separated, photovoltaic layer, comprising a conducting polymer and a fullerene derivative, is coated onto a thin metal wire. A second wire, coated with a silver film, serving as the counter electrode, is wrapped around the first wire. Both wires are encased in a transparent polymer cladding. Incident light is focused by the cladding onto to the photovoltaic layer even when it is completely shadowed by the counter electrode. Efficiency values of the wires range from 2.79% to 3.27%.

Aspects of forming metal-clad melt-processed Y-Ba-Cu-O tapes

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

Kozlowski, G.; Oberly, C.E.; Ho, J.

1991-03-01

This paper reports on melt-processing of Y-Ba-Cu-O superconductor in a usable form for magnet winding which requires the development of a cladding with demanding properties. Numerous recent efforts in cold forming Bi-based superconductor tapes have been successful because a silver tube can be used to constrain the ceramic material, which is sintered at much lower temperature than the Y-Ba-Cu-O. Typical high temperature metals which can be used to encase Y-Ba-Cu-O during sintering do not permit ready diffusion of oxygen as silver does. Recently, the full or partial recovery of superconductivity has been achieved in transition-metal- doped Y-Ba-Cu-O due to themore » partial-melt processing.« less

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2011 CFR

2011-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2010 CFR

2010-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2013 CFR

2013-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2012 CFR

2012-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

40 CFR 280.11 - Interim prohibition for deferred UST systems.

Code of Federal Regulations, 2014 CFR

2014-07-01

...) Will prevent releases due to corrosion or structural failure for the operational life of the UST system; (2) Is cathodically protected against corrosion, constructed of noncorrodible material, steel clad... substance. (b) Notwithstanding paragraph (a) of this section, an UST system without corrosion protection may...

Investigation of semiconductor clad optical waveguides

NASA Technical Reports Server (NTRS)

Batchman, T. E.; Mcwright, G. M.

1982-01-01

Glass waveguides are studied because of the ease and economy of fabricating devices in glass. All calculations are based on the assumption of a glass guide and substrate, but the effects being studied will occur on other materials if the proper refractive indices are used in the calculations.

Optical fiber evanescent absorption sensors for high-temperature gas sensing in advanced coal-fired power plants

NASA Astrophysics Data System (ADS)

Buric, Michael P.; Ohodnicky, Paul R.; Duy, Janice

2012-10-01

Modern advanced energy systems such as coal-fired power plants, gasifiers, or similar infrastructure present some of the most challenging harsh environments for sensors. The power industry would benefit from new, ultra-high temperature devices capable of surviving in hot and corrosive environments for embedded sensing at the highest value locations. For these applications, we are currently exploring optical fiber evanescent wave absorption spectroscopy (EWAS) based sensors consisting of high temperature core materials integrated with novel high temperature gas sensitive cladding materials. Mathematical simulations can be used to assist in sensor development efforts, and we describe a simulation code that assumes a single thick cladding layer with gas sensitive optical constants. Recent work has demonstrated that Au nanoparticle-incorporated metal oxides show a potentially useful response for high temperature optical gas sensing applications through the sensitivity of the localized surface plasmon resonance absorption peak to ambient atmospheric conditions. Hence, the simulation code has been applied to understand how such a response can be exploited in an optical fiber based EWAS sensor configuration. We demonstrate that interrogation can be used to optimize the sensing response in such materials.

Analysis of unclad and sub-clad semi-elliptical flaws in pressure vessel steels

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

Irizarry-Quinones, H.; Macdonald, B.D.; McAfee, W.J.

This study was conducted to support warm prestressing experiments on unclad and sub-clad flawed beams loaded in pure bending. Two cladding yield strengths were investigated: 0.6 Sy and 0.8 Sy, where Sy is the yield strength of the base metal. Cladding and base metal were assumed to be stress free at the stress relief temperature for the 3D elastic-plastic finite element analysis used to model the experiments. The model results indicated that when cooled from the stress relief temperature, the cladding was put in tension due to its greater coefficient of thermal expansion. When cooled, the cladding exhibited various amountsmore » of tensile yielding. The degree of yielding depended on the amount of cooling and the strength of the cladding relative to that of the base metal. When subjected to tensile bending stress, the sub-clad flaw elastic-plastic stress intensity factor, K{sub I}(J), was at first dominated by crack closing force due to tensile yielding in the cladding. Thus, imposed loads initially caused no increase in K{sub I}(J) near the clad-base interface. However, K{sub I}(J) at the flaw depth was little affected. When the cladding residual stress was overcome, K{sub I}(J) gradually increased until the cladding began to flow. Thereafter, the rate at which K{sub I}(J) increased with load was the same as that of an unclad beam. A plastic zone corrected K{sub I} approximation for the unclad flaw was found by the superposition of standard Newman and Raju solutions with those due to a cladding crack closure force approximated by the Kaya and Erdogan solution. These elastic estimates of the effect of cladding in reducing the crack driving force were quite in keeping with the 3D elastic-plastic finite element solution for the sub-clad flaw. The results were also compared with the analysis of clad beam experiments by Keeney and the conclusions by Miyazaki, et al. A number of sub-clad flaw specimens not subjected to warm prestressing were thought to have suffered degraded toughness caused by locally intensified strain aging embrittlement (LISAE) due to welding over the preexisting flaw.« less

An allowable cladding peak temperature for spent nuclear fuels in interim dry storage

NASA Astrophysics Data System (ADS)

Cha, Hyun-Jin; Jang, Ki-Nam; Kim, Kyu-Tae

2018-01-01

Allowable cladding peak temperatures for spent fuel cladding integrity in interim dry storage were investigated, considering hydride reorientation and mechanical property degradation behaviors of unirradiated and neutron irradiated Zr-Nb cladding tubes. Cladding tube specimens were heated up to various temperatures and then cooled down under tensile hoop stresses. Cool-down specimens indicate that higher heat-up temperature and larger tensile hoop stress generated larger radial hydride precipitation and smaller tensile strength and plastic hoop strain. Unirradiated specimens generated relatively larger radial hydride precipitation and plastic strain than did neutron irradiated specimens. Assuming a minimum plastic strain requirement of 5% for cladding integrity maintenance in interim dry storage, it is proposed that a cladding peak temperature during the interim dry storage is to keep below 250 °C if cladding tubes are cooled down to room temperature.

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

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

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

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

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

Bernander, O.; Haga, I.; Segerberg, F.

BS>From international nuclear industries fair; Basel, Switzerland (16 Oct 1972). Although the present status of the boiling water reactor is one of proven technology, design refinements and technical innovations are still being made to further improve reliability, economy and safety. The new standard ASEA- ATOM BWR features a number of such refinements and design improvements involving main circulation punips, containment design, refuelling system and off-gas treatment plant. In some respects the nuclear and hydraulic design of the ASEA- ATOM BWR differs from that adopted by other BWR manufacturers. Since the Oskarshamn I plant was the first nuclear power station havingmore » these features an extensive physics and hydraulics test program was made during the reactor start- up. The results of these tests have fully confirmed the ability of calculation methods to predict the behavior of the reactor. (auth)« less

Theoretical analysis of swelling characteristics of cylindrical uranium dioxide fuel pins with a niobium - 1-percent-zirconium clad

NASA Technical Reports Server (NTRS)

Saltsman, J. F.

1973-01-01

The relations between clad creep strain and fuel volume swelling are shown for cylindrical UO2 fuel pins with a Nb-1Zr clad. These relations were obtained by using the computer code CYGRO-2. These clad-strain - fuel-volume-swelling relations may be used with any fuel-volume-swelling model, provided the fuel volume swelling is isotropic and independent of the clad restraints. The effects of clad temperature (over a range from 118 to 1642 K (2010 to 2960 R)), pin diameter, clad thickness and central hole size in the fuel have been investigated. In all calculations the irradiation time was 500 hours. The burnup rate was varied.

Influence of laser radiation on structure and properties of steels and alloys

NASA Astrophysics Data System (ADS)

Tarasova, T.; Popova, E.

2013-03-01

In present study, and laser alloying of different steels and laser cladding of Ti and SiC powders mixtures was carried out, and microstructure, as well as microhardness profile and wear properties were examined. Research of the influence of lasers alloying modes on the elastic and plastic characteristics of the surface was conducted. As a result of chemical reactions in the cladded layer, a new phase (TiC) was synthesized during cladding process. The results showed that, in the clad layer, TiC was solidified to form dendrites in the clad zone. Produced coatings have high microhardness values in the upper and middle clad areas, about two time higher than clad matrix microhardness.

Measurement and removal of cladding light in high power fiber systems

NASA Astrophysics Data System (ADS)

Walbaum, Till; Liem, Andreas; Schreiber, Thomas; Eberhardt, Ramona; Tünnermann, Andreas

2018-02-01

The amount of cladding light is important to ensure longevity of high power fiber components. However, it is usually measured either by adding a cladding light stripper (and thus permanently modifying the fiber) or by using a pinhole to only transmit the core light (ignoring that there may be cladding mode content in the core area). We present a novel noninvasive method to measure the cladding light content in double-clad fibers based on extrapolation from a cladding region of constant average intensity. The method can be extended to general multi-layer radially symmetric fibers, e.g. to evaluate light content in refractive index pedestal structures. To effectively remove cladding light in high power systems, cladding light strippers are used. We show that the stripping efficiency can be significantly improved by bending the fiber in such a device and present respective experimental data. Measurements were performed with respect to the numerical aperture as well, showing the dependency of the CLS efficiency on the NA of the cladding light and implying that efficiency data cannot reliably be given for a certain fiber in general without regard to the properties of the guided light.

Electroslag Strip Cladding of Steam Generators With Alloy 690

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

Consonni, M.; Maggioni, F.; Brioschi, F.

2006-07-01

The present paper details the results of electroslag cladding and tube-to-tubesheet welding qualification tests conducted by Ansaldo-Camozzi ESC with Alloy 690 (Alloy 52 filler metal) on steel for nuclear power stations' steam generators shell, tubesheet and head; the possibility of submerged arc cladding on first layer was also considered. Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layermore » leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallization effect of second cladding layer. Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding. (authors)« less

Building America Case Study: Initial and Long-Term Movement of Cladding Installed Over Exterior Rigid Insulation (Fact Sheet)

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

Not Available

Changes in the International Energy Conservation Code (IECC) from 2009 to 2012 have resulted in the use of exterior rigid insulation becoming part of the prescriptive code requirements. With more jurisdictions adopting the 2012 IECC builders are going to finding themselves required to incorporate exterior insulation in the construction of their exterior wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location. However, there has beenmore » a significant resistance to its widespread implementation due to a lack of research and understanding of the mechanisms involved and potential creep effects of the assembly under the sustained dead load of a cladding. This research was an extension on previous research conducted by BSC in 2011, and 2012. Each year the understanding of the system discrete load component interactions, as well as impacts of environmental loading has increased. The focus of the research was to examine more closely the impacts of screw fastener bending on the total system capacity, effects of thermal expansion and contraction of materials on the compressive forces in the assembly, as well as to analyze a full years worth of cladding movement data from assemblies constructed in an exposed outdoor environment.« less

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Orientation-Dependent Displacement Sensor Using an Inner Cladding Fiber Bragg Grating.

PubMed

Yang, Tingting; Qiao, Xueguang; Rong, Qiangzhou; Bao, Weijia

2016-09-11

An orientation-dependent displacement sensor based on grating inscription over a fiber core and inner cladding has been demonstrated. The device comprises a short piece of multi-cladding fiber sandwiched between two standard single-mode fibers (SMFs). The grating structure is fabricated by a femtosecond laser side-illumination technique. Two well-defined resonances are achieved by the downstream both core and cladding fiber Bragg gratings (FBGs). The cladding resonance presents fiber bending dependence, together with a strong orientation dependence because of asymmetrical distribution of the "cladding" FBG along the fiber cross-section.

24 CFR 3280.808 - Wiring methods and materials.

Code of Federal Regulations, 2012 CFR

2012-04-01

... the appliance terminal connections to a readily accessible outlet box placed at least one foot from... manufactured homes. (b) Nonmetallic outlet boxes shall be acceptable only with nonmetallic cable. (c... not exceeding 41/2 feet and within 12 inches from every cabinet, box or fitting. (e) Metal-clad and...

24 CFR 3280.808 - Wiring methods and materials.

Code of Federal Regulations, 2013 CFR

2013-04-01

... the appliance terminal connections to a readily accessible outlet box placed at least one foot from... manufactured homes. (b) Nonmetallic outlet boxes shall be acceptable only with nonmetallic cable. (c... not exceeding 41/2 feet and within 12 inches from every cabinet, box or fitting. (e) Metal-clad and...

The effect of laser process parameters on microstructure and dilution rate of cladding coatings

NASA Astrophysics Data System (ADS)

Bin, Liu; Heping, Liu; Xingbin, Jing; Yuxin, Li; Peikang, Bai

2018-02-01

In order to broaden the range of application of Q235 steel, it is necessary to repair the surface of steel. High performance 316L stainless steel coating was successfully obtained on Q235 steel by laser cladding technology. The effect of laser cladding parameters on the geometrical size and appearance of single cladding layer was investigated. The experimental results show that laser current has an important influence on the surface morphology of single channel cladding. When the current is from 155A to 165A, the cladding coating becomes smooth. The laser current has an effect on the geometric cross section size and dilution rate of single cladding. The results revealed that with the rising of laser current, the width, height and depth of layer increase gradually. With the rising of laser current, the dilution rate of cladding layer is gradually increasing.

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Use of Seawater for Air Conditioning at Waikiki Convention Center

DTIC Science & Technology

1994-01-01

construct heat exchangers of either titanium or zinc-clad aluminum . It is likely that these could be ordered "off the shelf*. Bio-fouling would not be a...Considerations J. Fouling K. Alternative Design: Directional Drilling System Design A. Seawater Heat Exchangers 1. Materials 2. Bio-fouling 3. Design... Heat exchangers are available in a variety of types, sizes and materials . There are standard models available by many manufactures, however, a single

Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

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

Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

2015-12-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are keptmore » open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

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

Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement formore » extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Cyclic crack growth behavior of reactor pressure vessel steels in light water reactor environments

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

Van Der Sluys, W.A.; Emanuelson, R.H.

1986-01-01

During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533b-1 steels in simulated 550/sup 0/F boiling water reactor (BWR) and 550/sup 0/F pressurized water reactor (PWR) environments. Areas investigated over the coursemore » of the test program included the effects of loading frequency and r ratio (Kmin-Kmax) on crack growth rate as a function of the stress intensity factor (deltaK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by deltaK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were in a 550/sup 0/F simulated BWR environment than in a 550/sup 0/F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency.« less

Formation of anomalous eutectic in Ni-Sn alloy by laser cladding

NASA Astrophysics Data System (ADS)

Wang, Zhitai; Lin, Xin; Cao, Yongqing; Liu, Fencheng; Huang, Weidong

2018-02-01

Ni-Sn anomalous eutectic is obtained by single track laser cladding with the scanning velocity from 1 mm/s to 10 mm/s using the Ni-32.5 wt.%Sn eutectic powders. The microstructure of the cladding layer and the grain orientations of anomalous eutectic were investigated. It is found that the microstructure is transformed from primary α-Ni dendrites and the interdendritic (α-Ni + Ni3Sn) eutectic at the bottom of the cladding layer to α-Ni and β-Ni3Sn anomalous eutectic at the top of the cladding layer, whether for single layer or multilayer laser cladding. The EBSD maps and pole figures indicate that the spatially structure of α-Ni phase is discontinuous and the Ni3Sn phase is continuous in anomalous eutectic. The transformation from epitaxial growth columnar at bottom of cladding layer to free nucleation equiaxed at the top occurs, i.e., the columnar to equiaxed transition (CET) at the top of cladding layer during laser cladding processing leads to the generation of anomalous eutectic.

Air ingression calculations for selected plant transients using MELCOR

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

Kmetyk, L.N.

1994-01-01

Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition tomore » the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression.« less

Multi-clad black display panel

DOEpatents

Veligdan, James T.; Biscardi, Cyrus; Brewster, Calvin

2002-01-01

A multi-clad black display panel, and a method of making a multi-clad black display panel, are disclosed, wherein a plurality of waveguides, each of which includes a light-transmissive core placed between an opposing pair of transparent cladding layers and a black layer disposed between transparent cladding layers, are stacked together and sawed at an angle to produce a wedge-shaped optical panel having an inlet face and an outlet face.

Research on microstructure properties of the TiC/Ni-Fe-Al coating prepared by laser cladding technology

NASA Astrophysics Data System (ADS)

Jiao, Junke; Xu, Zifa; Zan, Shaoping; Zhang, Wenwu; Sheng, Liyuan

2017-10-01

In this paper, the laser cladding method was used to preparation the TiC reinforced Ni-Fe-Al coating on the Ni base superalloy. The Ti/Ni-Fe-Al powder was preset on the Ni base superalloy and the powder layer thickness is 0.5mm. A fiber laser was used the melting Ti/Ni-Fe-Al powder in an inert gas environment. The shape of the cladding layer was tested using laser scanning confocal microscope (LSCM) under different cladding parameters such as the laser power, the melting velocity and the defocused amount. The microstructure, the micro-hardness was tested by LSCM, SEM, Vickers hardness tester. The test result showed that the TiC particles was distributed uniformly in the cladding layer and hardness of the cladding layer was improved from 180HV to 320HV compared with the Ni-Fe-Al cladding layer without TiC powder reinforced, and a metallurgical bonding was produced between the cladding layer and the base metal. The TiC powder could make the Ni-Fe-Al cladding layer grain refining, and the more TiC powder added in the Ni-Fe-Al powder, the smaller grain size was in the cladding layer.

Laser Surface Treatment of Sintered Alumina

NASA Astrophysics Data System (ADS)

Hagemann, R.; Noelke, C.; Kaierle, S.; Wesling, V.

Sintered alumina ceramics are used as refractory materials for industrial aluminum furnaces. In this environment the ceramic surface is in permanent contact with molten aluminum resulting in deposition of oxidic material on its surface. Consequently, a lower volume capacity as well as thermal efficiency of the furnaces follows. To reduce oxidic adherence of the ceramic material, two laser-based surface treatment processes were investigated: a powder- based single-step laser cladding and a laser surface remelting. Main objective is to achieve an improved surface quality of the ceramic material considering the industrial requirements as a high process speed.

Real time monitoring of environmental crack growth in BWRs

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

Hale, D.; Diehl, C.G.

1988-01-01

A comprehensive field test program was recently completed at several Boiling Water Reactors (BWR) to quantify the effect of coolant impurities on the initiation and growth of stress corrosion cracks. A new technology was utilized which allows for real time monitoring of stress corrosion crack growth rates. The BWR environments were characterized using Ion Chromatography and Electro Chemical Potential (ECP) measurements. The effects of typical water chemistry transients and startups were quantified.

The Longitudinal Force Measurement of CWR Tracks with Hetero-Cladding FBG Sensors: A Proof of Concept.

PubMed

Shao, Li-Yang; Zhang, Meng; Xie, Kaize; Zhang, Xinpu; Wang, Ping; Yan, Lianshan

2016-12-18

A new method has been proposed to accurately determine longitudinal additional force in continuous welded rail (CWR) on bridges via hetero-cladding fiber Bragg grating (HC-FBG) sensors. The HC-FBG sensor consists of two FBGs written in the same type of fiber but with different cladding diameters. The HC-FBGs have the same temperature sensitivity but different strain sensitivity because of the different areas of the cross section. The differential strain coefficient is defined as the relative wavelength differences of two FBGs with the change of applied longitudinal force. In the verification experiment in the lab, the HC-FBGs were attached on a section of rail model of which the material property is the same as that of rail on line. The temperature and differential strain sensitivity were calibrated using a universal testing machine. As shown by the test results, the linearity between the relative wavelength difference and the longitudinal additional force is greater than 0.9999. The differential strain sensitivity is 4.85 × 10 -6 /N. Moreover, the relative wavelength difference is not affected by the temperature change. Compared to the theoretical results, the accumulated error is controlled within 5.0%.

Initial and Long-Term Movement of Cladding Installed Over Exterior Rigid Insulation

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

Baker, Peter

Changes in the International Energy Conservation Code (IECC) from 2009 to 2012 have resulted in the use of exterior rigid insulation becoming part of the prescriptive code requirements. With more jurisdictions adopting the 2012 IECC builders will be required to incorporate exterior insulation in the construction of their exterior wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use of wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location. This research was an extension onmore » previous research conducted by Building Science Corporation in 2011, and 2012. Each year the understanding of the system discrete load component interactions, as well as impacts of environmental loading, has increased. The focus of the research was to examine more closely the impacts of screw fastener bending on the total system capacity, effects of thermal expansion and contraction of materials on the compressive forces in the assembly, as well as to analyze a full year’s worth of cladding movement data from assemblies constructed in an exposed outdoor environment.« less

Uncertainty budgets for liquid waveguide CDOM absorption measurements.

PubMed

Lefering, Ina; Röttgers, Rüdiger; Utschig, Christian; McKee, David

2017-08-01

Long path length liquid waveguide capillary cell (LWCC) systems using simple spectrometers to determine the spectral absorption by colored dissolved organic matter (CDOM) have previously been shown to have better measurement sensitivity compared to high-end spectrophotometers using 10 cm cuvettes. Information on the magnitude of measurement uncertainties for LWCC systems, however, has remained scarce. Cross-comparison of three different LWCC systems with three different path lengths (50, 100, and 250 cm) and two different cladding materials enabled quantification of measurement precision and accuracy, revealing strong wavelength dependency in both parameters. Stable pumping of the sample through the capillary cell was found to improve measurement precision over measurements made with the sample kept stationary. Results from the 50 and 100 cm LWCC systems, with higher refractive index cladding, showed systematic artifacts including small but unphysical negative offsets and high-frequency spectral perturbations due to limited performance of the salinity correction. In comparison, the newer 250 cm LWCC with lower refractive index cladding returned small positive offsets that may be physically correct. After null correction of measurements at 700 nm, overall agreement of CDOM absorption data at 440 nm was found to be within 5% root mean square percentage error.

A study of TiB2/TiB gradient coating by laser cladding on titanium alloy

NASA Astrophysics Data System (ADS)

Lin, Yinghua; Lei, Yongping; Li, Xueqiao; Zhi, Xiaohui; Fu, Hanguang

2016-07-01

TiB2/TiB gradient coating has been fabricated by a laser cladding technique on the surface of a Ti-6Al-4V substrate using TiB2 powder as the cladding material. The microstructure and mechanical properties of the gradient coating were analyzed by SEM, EPMA, XRD, TEM and an instrument to measure hardness. With the increasing distance from the coating surface, the content of TiB2 particles gradually decreased, but the content of TiB short fibers gradually increased. Meanwhile, the micro-hardness and the elastic modulus of the TiB2/TiB coating showed a gradient decreasing trend, but the fracture toughness showed a gradient increasing trend. The fracture toughness of the TiB2/TiB coating between the center and the bottom was improved, primarily due to the debonding of TiB2 particles and the high fracture of TiB short fibers, and the fracture position of TiB short fiber can be moved to an adjacent position. However, the debonding of TiB2 particles was difficult to achieve at the surface of the TiB2/TiB coating.

Single Mode SU8 Polymer Based Mach-Zehnder Interferometer for Bio-Sensing Application

NASA Astrophysics Data System (ADS)

Boiragi, Indrajit; Kundu, Sushanta; Makkar, Roshan; Chalapathi, Krishnamurthy

2011-10-01

This paper explains the influence of different parameters to the sensitivity of an optical waveguide Mach-Zehnder Interferometer (MZI) for real time detection of biomolecules. The sensing principle is based on the interaction of evanescence field with the biomolecules that get immobilized on sensing arm. The sensitivity has been calculated by varying the sensing window length, wavelength and concentration of bio-analyte. The maximum attainable sensitivity for the preferred design is the order of 10-8 RIU at 840 nm wavelength with a sensing window length of 1cm. All the simulation work has been carried out with Opti-BPMCAD for the optimization of MZI device parameters. The SU8 polymers are used as a core and clad material to fabricate the waveguide. The refractive index of cladding layer is optimized by varying the curing temperature for a fixed time period and the achieved index difference between core and clad is Δn = 0.0151. The fabricated MZI device has been characterized with LASER beam profiler at 840 nm wavelength. This study demonstrates the effectiveness of the different parameter to the sensitivity of a single mode optical waveguide Mach-Zehnder Interferometer for bio-sensing application.

Thermally Stable Siloxane Hybrid Matrix with Low Dielectric Loss for Copper-Clad Laminates for High-Frequency Applications.

PubMed

Kim, Yong Ho; Lim, Young-Woo; Kim, Yun Hyeok; Bae, Byeong-Soo

2016-04-06

We report vinyl-phenyl siloxane hybrid material (VPH) that can be used as a matrix for copper-clad laminates (CCLs) for high-frequency applications. The CCLs, with a VPH matrix fabricated via radical polymerization of resin blend consisting of sol-gel-derived linear vinyl oligosiloxane and bulky siloxane monomer, phenyltris(trimethylsiloxy)silane, achieve low dielectric constant (Dk) and dissipation factor (Df). The CCLs with the VPH matrix exhibit excellent dielectric performance (Dk = 2.75, Df = 0.0015 at 1 GHz) with stability in wide frequency range (1 MHz to 10 GHz) and at high temperature (up to 275 °C). Also, the VPH shows good flame resistance without any additives. These results suggest the potential of the VPH for use in high-speed IC boards.

TRIGA MARK-II source term

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

Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences ofmore » results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.« less

Strengthening Effect of Incremental Shear Deformation on Ti Alloy Clad Plate with a Ni-Based Alloy Laser-Clad Layer

NASA Astrophysics Data System (ADS)

Zhao, W.; Zha, G. C.; Kong, F. X.; Wu, M. L.; Feng, X.; Gao, S. Y.

2017-05-01

A Ti-6Al-4V alloy clad plate with a Tribaloy 700 alloy laser-clad layer is subjected to incremental shear deformation, and we evaluate the structural evolution and mechanical properties of the specimens. Results indicate the significance of the incremental shear deformation on the strengthening effect. The wear resistance and Vickers hardness of the laser-clad layer are enhanced due to increased dislocation density. The incremental shear deformation can increase the bonding strength of the laser-clad layer and the corresponding substrate and can break the columnar crystals in the laser-clad layer near the interface. These phenomena suggest that shear deformation eliminates the defects on the interface of the laser-clad layer and the substrate. Substrate hardness is evidently improved, and the strengthening effect is caused by the increased dislocation density and shear deformation. This deformation can then transform the α- and β-phases in the substrate into a high-intensity ω-phase.

Orientation-dependent fiber-optic accelerometer based on grating inscription over fiber cladding.

PubMed

Rong, Qiangzhou; Qiao, Xueguang; Guo, Tuan; Bao, Weijia; Su, Dan; Yang, Hangzhou

2014-12-01

An orientation-sensitive fiber-optic accelerometer based on grating inscription over fiber cladding has been demonstrated. The sensor probe comprises a compact structure in which a short section of thin-core fiber (TCF) stub containing a "cladding" fiber Bragg grating (FBG) is spliced to another single-mode fiber (SMF) without any lateral offset. A femtosecond laser side-illumination technique was utilized to ensure that the grating inscription remains close to the core-cladding interface of the TCF. The core mode and the cladding mode of the TCF are coupled at the core-mismatch junction, and two well-defined resonances in reflection appear from the downstream FBG, in which the cladding resonance exhibits a strong polarization and bending dependence due to the asymmetrical distribution of the cladding FBG along the fiber cross section. Strong orientation dependence of the vibration (acceleration) measurement has been achieved by power detection of the cladding resonance. Meanwhile, the unwanted power fluctuations and temperature perturbations can be referenced out by monitoring the fundamental core resonance.

Laser Cladding for Crack Repair of CMSX-4 Single-Crystalline Turbine Parts

NASA Astrophysics Data System (ADS)

Rottwinkel, Boris; Nölke, Christian; Kaierle, Stefan; Wesling, Volker

2017-03-01

The increase of the lifetime of modern single crystalline (SX) turbine blades is of high economic priority. The currently available repair methods using polycrystalline cladding of the damaged area do not address the issue of monocrystallinity and are restricted to few areas of the blade. The tip area of the blade is most prone to damage and undergoes the most wear, erosion and cracking during its lifetime. To repair such defects, the common procedure is to remove the whole tip with the damaged area and rebuild it by applying a polycrystalline solidification of the material. The repair of small cracks is conducted in the same way. To reduce repair cost, the investigation of a manufacturing process to repair these cracked areas while maintaining single-crystal solidification is of high interest as this does not diminish material properties and thereby its lifetime. To establish this single-crystal solidification, the realization of a directed temperature gradient is needed. The initial scope of this work is the computational prediction of the temperature field that arises and its verification during the process. The laser cladding process of CMSX-4 substrates was simulated and the necessary parameters calculated. These parameters were then applied to notched substrates and their microstructures analyzed. Starting with a simulation of the temperature field using ANSYS®, a process to repair parts of single crystalline nickel-based alloys was developed. It could be shown that damages to the tip area and cracks can be repaired by establishing a specific temperature gradient during the repair process in order to control the solidification process.

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

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

Bragg-Sitton, Shannon Michelle

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, parametric studies, and selection of system codes. Themore » Cladding and Core Materials and Fuel Concepts task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment Task Force is chaired by Shannon Bragg-Sitton (INL), while the Cladding Task Force will be chaired by a representative from France (Marie Moatti, Electricite de France [EdF]) and the Fuels Task Force will be chaired by a representative from Japan (Masaki Kurata, Japan Atomic Energy Agency [JAEA]). This report provides an overview of the Systems Assessment Task Force charter and status of work accomplishment.« less

Evolution of transmission spectra of double cladding fiber during etching

NASA Astrophysics Data System (ADS)

Ivanov, Oleg V.; Tian, Fei; Du, Henry

2017-11-01

We investigate the evolution of optical transmission through a double cladding fiber-optic structure during etching. The structure is formed by a section of SM630 fiber with inner depressed cladding between standard SMF-28 fibers. Its transmission spectrum exhibits two resonance dips at wavelengths where two cladding modes have almost equal propagation constants. We measure transmission spectra with decreasing thickness of the cladding and show that the resonance dips shift to shorter wavelengths, while new dips of lower order modes appear from long wavelength side. We calculate propagation constants of cladding modes and resonance wavelengths, which we compare with the experiment.

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

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

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

Capturing reflected cladding modes from a fiber Bragg grating with a double-clad fiber coupler.

PubMed

Baiad, Mohamad Diaa; Gagné, Mathieu; Lemire-Renaud, Simon; De Montigny, Etienne; Madore, Wendy-Julie; Godbout, Nicolas; Boudoux, Caroline; Kashyap, Raman

2013-03-25

We present a novel measurement scheme using a double-clad fiber coupler (DCFC) and a fiber Bragg grating (FBG) to resolve cladding modes. Direct measurement of the optical spectra and power in the cladding modes is obtained through the use of a specially designed DCFC spliced to a highly reflective FBG written into slightly etched standard photosensitive single mode fiber to match the inner cladding diameter of the DCFC. The DCFC is made by tapering and fusing two double-clad fibers (DCF) together. The device is capable of capturing backward propagating low and high order cladding modes simply and efficiently. Also, we demonstrate the capability of such a device to measure the surrounding refractive index (SRI) with an extremely high sensitivity of 69.769 ± 0.035 μW/RIU and a resolution of 1.433 × 10(-5) ± 8 × 10(-9) RIU between 1.37 and 1.45 RIU. The device provides a large SRI operating range from 1.30 to 1.45 RIU with sufficient discrimination for all individual captured cladding modes. The proposed scheme can be adapted to many different types of bend, temperature, refractive index and other evanescent wave based sensors.

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

DOE PAGES

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.; ...

2016-03-16

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

TE and TM guided modes in an air waveguide with negative-index-material cladding.

PubMed

D'Aguanno, G; Mattiucci, N; Scalora, M; Bloemer, M J

2005-04-01

We numerically demonstrate that a planar waveguide in which the inner layer is a gas with refractive index n0 = 1, sandwiched between two identical semi-infinite layers of a negative index material, can support both transverse electric and transverse magnetic guided modes with low losses. Recent developments in the design of metamaterials with an effective negative index suggest that this waveguide could operate in the infrared region of the spectrum.

Lithographically-Scribed Planar Holographic Optical CDMA Devices and Systems

DTIC Science & Technology

2007-02-15

operate with quite high refractive index contrast (order 0.5). Thin -filn filter devices are viewed as relatively low in chromatic dispersion. We have...stack consists of planar interfaces between materials of refractive index n, and n,. Let An = In2 - nil and n = (n, - n1)/2. The planar interfaces are... index ). It may be desirable to have a relatively large refractive index differential when diffractive elements are formed from cladding material at a

Test Results of Heat Exchanger Cleaning in Support of Ocean Thermal Energy Conversion.

DTIC Science & Technology

1980-12-01

tests evaluated the performance of three in-situ cleaning techniques in two potential heat exchanger materials ...1-6. 41Mann, M. J., 1979, "Possible Cu-Ni-Clad Steel Material and Abrasive Slurry Cleaning System for Plate-Fin-Type OTEC Heat Exchangers ," in...of a Shell-less Folded Aluminum Tube, OTEC Heat Exchanger ," Proceedings of the Sixth OTEC Conference, Washington, DC, June 19-22, 1978, pp 12.8-1

Transmission of laser pulses with high output beam quality using step-index fibers having large cladding

DOEpatents

Yalin, Azer P; Joshi, Sachin

2014-06-03

An apparatus and method for transmission of laser pulses with high output beam quality using large core step-index silica optical fibers having thick cladding, are described. The thick cladding suppresses diffusion of modal power to higher order modes at the core-cladding interface, thereby enabling higher beam quality, M.sup.2, than are observed for large core, thin cladding optical fibers. For a given NA and core size, the thicker the cladding, the better the output beam quality. Mode coupling coefficients, D, has been found to scale approximately as the inverse square of the cladding dimension and the inverse square root of the wavelength. Output from a 2 m long silica optical fiber having a 100 .mu.m core and a 660 .mu.m cladding was found to be close to single mode, with an M.sup.2=1.6. Another thick cladding fiber (400 .mu.m core and 720 .mu.m clad) was used to transmit 1064 nm pulses of nanosecond duration with high beam quality to form gas sparks at the focused output (focused intensity of >100 GW/cm.sup.2), wherein the energy in the core was <6 mJ, and the duration of the laser pulses was about 6 ns. Extending the pulse duration provided the ability to increase the delivered pulse energy (>20 mJ delivered for 50 ns pulses) without damaging the silica fiber.

Robust cladding light stripper for high-power fiber lasers using soft metals.

PubMed

Babazadeh, Amin; Nasirabad, Reza Rezaei; Norouzey, Ahmad; Hejaz, Kamran; Poozesh, Reza; Heidariazar, Amir; Golshan, Ali Hamedani; Roohforouz, Ali; Jafari, S Naser Tabatabaei; Lafouti, Majid

2014-04-20

In this paper we present a novel method to reliably strip the unwanted cladding light in high-power fiber lasers. Soft metals are utilized to fabricate a high-power cladding light stripper (CLS). The capability of indium (In), aluminum (Al), tin (Sn), and gold (Au) in extracting unwanted cladding light is examined. The experiments show that these metals have the right features for stripping the unwanted light out of the cladding. We also find that the metal-cladding contact area is of great importance because it determines the attenuation and the thermal load on the CLS. These metals are examined in different forms to optimize the contact area to have the highest possible attenuation and avoid localized heating. The results show that sheets of indium are very effective in stripping unwanted cladding light.

24 CFR 3280.808 - Wiring methods and materials.

Code of Federal Regulations, 2014 CFR

2014-04-01

... Electrical Code, NFPA No. 70-2005, must be used in manufactured homes. (b) Nonmetallic outlet boxes shall be... every cabinet, box or fitting. (e) Metal-clad and nonmetallic cables shall be permitted to pass through... junction box or range or dryer receptacle. (h) Threaded rigid metal conduit shall be provided with a...

Lightweight long life heat exchanger

NASA Technical Reports Server (NTRS)

Moore, E. K.

1975-01-01

The design, fabrication, and evaluation of a full scale shuttle-type condensing heat exchanger constructed of aluminum and utilizing aluminum clad titanium parting sheets is described. A long term salt spray test of candidate parting sheet specimens is described. The results of an investigation into an alternate method of making composite sheet material are discussed.

Studies of thermionic materials for space power applications

NASA Technical Reports Server (NTRS)

1971-01-01

Service life tests of LC-8 and LC-9 carbide-fueled thermionic converters are discussed. Post operational tests of the converters to show emitter diametric change, microstructures of cladding and fuel, and analysis of fuel composition are described. The fabrication and performance of high temperature thermocouples used in the test procedures are included.

Investigation of Experimental Lightweight Firewall Materials for A/C Engine Bay Applications.

DTIC Science & Technology

1985-04-01

umidity conditions. The metal clad Johns - Manville samples were basically qual. They do not provide a vapor barrier in the configurations tested, but an be...made to by the use of coatings or backings. The flexibility of the 3M nd Johns - Manville samples make them excellent choices for subsystem fire

Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

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

Was, Gary

This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

Use of eddy current mixes to solve a weld examination application

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

Ward, R.C.; LaBoissonniere, A.

1995-12-31

The augmentation of typical nondestructive (i.e., ultrasound) weld inspection techniques by the use of eddy current tools may significantly enhance the quality and reliability of weld inspections. One recent example is the development of an eddy current technique for use in the examination of BWR core shroud welds, where multi-frequency mixes are used to eliminate signals coming from the weld material so that the examination of the heat affected zone is enhanced. An analysis tool most commonly associated with ultrasound examinations, the C-Scan based on gated information, may be implemented with eddy current data to enhance analysis.

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Ion irradiation testing and characterization of FeCrAl candidate alloys

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

Anderoglu, Osman; Aydogan, Eda; Maloy, Stuart Andrew

2014-10-29

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commerciallymore » available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.« less

High-speed electro-optic switch based on nonlinear polymer-clad waveguide incorporated with quasi-in-plane coplanar waveguide electrodes

NASA Astrophysics Data System (ADS)

Jiang, Ming-Hui; Wang, Xi-Bin; Xu, Qiang; Li, Ming; Niu, Dong-Hai; Sun, Xiao-Qiang; Wang, Fei; Li, Zhi-Yong; Zhang, Da-Ming

2018-01-01

Nonlinear optical (NLO) polymer is a promising material for active waveguide devices that can provide large bandwidth and high-speed response time. However, the performance of the active devices is not only related to the waveguide materials, but also related to the waveguide and electrode structures. In this paper, a high-speed Mach-Zehnder interferometer (MZI) type of electro-optic (EO) switch based on NLO polymer-clad waveguide was fabricated. The quasi-in-plane coplanar waveguide electrodes were also introduced to enhance the poling and modulating efficiency. The characteristic parameters of the waveguide and electrode were carefully designed and simulated. The switches were fabricated by the conventional micro-fabrication process. Under 1550-nm operating wavelength, a typical fabricated switch showed a low insertion loss of 10.2 dB, and the switching rise time and fall time were 55.58 and 57.98 ns, respectively. The proposed waveguide and electrode structures could be developed into other active EO devices and also used as the component in the polymer-based large-scale photonic integrated circuit.

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

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

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Modeling the Formation of Transverse Weld during Billet-on-Billet Extrusion

PubMed Central

Mahmoodkhani, Yahya; Wells, Mary; Parson, Nick; Jowett, Chris; Poole, Warren

2014-01-01

A comprehensive mathematical model of the hot extrusion process for aluminum alloys has been developed and validated. The plasticity module was developed using a commercial finite element package, DEFORM-2D, a transient Lagrangian model which couples the thermal and deformation phenomena. Validation of the model against industrial data indicated that it gave excellent predictions of the pressure during extrusion. The finite element predictions of the velocity fields were post-processed to calculate the thickness of the surface cladding as one billet is fed in after another through the die (i.e., the transverse weld). The mathematical model was then used to assess the effect a change in feeder dimensions would have on the shape, thickness and extent of the transverse weld during extrusion. Experimental measurements for different combinations of billet materials show that the model is able to accurately predict the transverse weld shape as well as the clad surface layer to thicknesses of 50 μm. The transverse weld is significantly affected by the feeder geometry shape, but the effects of ram speed, billet material and temperature on the transverse weld dimensions are negligible. PMID:28788629

Reducing tool wear by partial cladding of critical zones in hot form tool by laser metal deposition

NASA Astrophysics Data System (ADS)

Vollmer, Robert; Sommitsch, Christof

2017-10-01

This paper points out a production method to reduce tool wear in hot stamping applications. Usually tool wear can be observed at locally strongly stressed areas superimposed with gliding movement between blank and tool surface. The shown solution is based on a partial laser cladding of the tool surface with a wear resistant coating to increase the lifespan of tool inserts. Preliminary studies showed good results applying a material combination of tungsten carbide particles embedded in a metallic matrix. Different Nickel based alloys welded on hot work tool steel (1.2343) were tested mechanically in the interface zone. The material with the best bonding characteristic is chosen and reinforced with spherical tungsten carbide particles in a second laser welding step. Since the machining of tungsten carbides is very elaborate a special manufacturing strategy is developed to reduce the milling effort as much as possible. On special test specimens milling tests are carried out to proof the machinability. As outlook a tool insert of a b-pillar is coated to perform real hot forming tests.

Impact of forced vital capacity loss on survival after the onset of chronic lung allograft dysfunction.

PubMed

Todd, Jamie L; Jain, Rahil; Pavlisko, Elizabeth N; Finlen Copeland, C Ashley; Reynolds, John M; Snyder, Laurie D; Palmer, Scott M

2014-01-15

Emerging evidence suggests a restrictive phenotype of chronic lung allograft dysfunction (CLAD) exists; however, the optimal approach to its diagnosis and clinical significance is uncertain. To evaluate the hypothesis that spirometric indices more suggestive of a restrictive ventilatory defect, such as loss of FVC, identify patients with distinct clinical, radiographic, and pathologic features, including worse survival. Retrospective, single-center analysis of 566 consecutive first bilateral lung recipients transplanted over a 12-year period. A total of 216 patients developed CLAD during follow-up. CLAD was categorized at its onset into discrete physiologic groups based on spirometric criteria. Imaging and histologic studies were reviewed when available. Survival after CLAD diagnosis was assessed using Kaplan-Meier and Cox proportional hazards models. Among patients with CLAD, 30% demonstrated an FVC decrement at its onset. These patients were more likely to be female, have radiographic alveolar or interstitial changes, and histologic findings of interstitial fibrosis. Patients with FVC decline at CLAD onset had significantly worse survival after CLAD when compared with those with preserved FVC (P < 0.0001; 3-yr survival estimates 9% vs. 48%, respectively). The deleterious impact of CLAD accompanied by FVC loss on post-CLAD survival persisted in a multivariable model including baseline demographic and clinical factors (P < 0.0001; adjusted hazard ratio, 2.73; 95% confidence interval, 1.86-4.04). At CLAD onset, a subset of patients demonstrating physiology more suggestive of restriction experience worse clinical outcomes. Further study of the biologic mechanisms underlying CLAD phenotypes is critical to improving long-term survival after lung transplantation.

Advanced Neutronics Tools for BWR Design Calculations

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

Santamarina, A.; Hfaiedh, N.; Letellier, R.

2006-07-01

This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007more » BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)« less

Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

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

Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.

2015-01-01

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational datamore » available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.« less

Evaluation of Two CEDA Weatherization Pilot Implementations of an Exterior Insulation and Over-Clad Retrofit Strategy for Residential Masonry Buildings in Chicago

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

Neuhauser, K.

This project examines the implementation of an exterior insulation and over-clad strategy for brick masonry buildings in Chicago. The strategy was implemented at a free-standing two story two-family dwelling and a larger free-standing multifamily building. The test homes selected for this research represent predominant housing types for the Chicago area. High heating energy use typical in these buildings threaten housing affordability. Uninsulated mass masonry wall assemblies also have a strongly detrimental impact on comfort. Significant changes to the performance of masonry wall assemblies is generally beyond the reach of typical weatherization (Wx) program resources. The Community and Economic Development Associationmore » of Cook County, Inc. (CEDA) has secured a Sustainable Energy Resources for Consumers (SERC) innovation grant sponsored by the United States Department of Energy (DOE). This grant provides CEDA the opportunity to pursue a pilot implementation of innovative approaches to retrofit in masonry wall enclosures. The exterior insulation and over-clad strategy implemented through this project was designed to allow implementation by contractors active in CEDA weatherization programs and using materials and methods familiar to these contractors. The retrofit measures are evaluated in terms of feasibility, cost and performance. Through observations of the strategies implemented, the research described in this report identifies measures critical to performance as well as conditions for wider adoption. The research also identifies common factors that must be considered in determining whether the exterior insulation and over-clad strategy is appropriate for the building.« less

CXCL4 Contributes to the Pathogenesis of Chronic Liver Allograft Dysfunction

PubMed Central

Li, Jing; Shi, Yuan; Xie, Ke-Liang; Yin, Hai-Fang; Yan, Lu-nan; Lau, Wan-yee; Wang, Guo-Lin

2016-01-01

Chronic liver allograft dysfunction (CLAD) remains the most common cause of patient morbidity and allograft loss in liver transplant patients. However, the pathogenesis of CLAD has not been completely elucidated. By establishing rat CLAD models, in this study, we identified the informative CLAD-associated genes using isobaric tags for relative and absolute quantification (iTRAQ) proteomics analysis and validated these results in recipient rat liver allografts. CXCL4, CXCR3, EGFR, JAK2, STAT3, and Collagen IV were associated with CLAD pathogenesis. We validated that CXCL4 is upstream of these informative genes in the isolated hepatic stellate cells (HSC). Blocking CXCL4 protects against CLAD by reducing liver fibrosis. Therefore, our results indicated that therapeutic approaches that neutralize CXCL4, a newly identified target of fibrosis, may represent a novel strategy for preventing and treating CLAD after liver transplantation. PMID:28053995

CXCL4 Contributes to the Pathogenesis of Chronic Liver Allograft Dysfunction.

PubMed

Li, Jing; Liu, Bin; Shi, Yuan; Xie, Ke-Liang; Yin, Hai-Fang; Yan, Lu-Nan; Lau, Wan-Yee; Wang, Guo-Lin

2016-01-01

Chronic liver allograft dysfunction (CLAD) remains the most common cause of patient morbidity and allograft loss in liver transplant patients. However, the pathogenesis of CLAD has not been completely elucidated. By establishing rat CLAD models, in this study, we identified the informative CLAD-associated genes using isobaric tags for relative and absolute quantification (iTRAQ) proteomics analysis and validated these results in recipient rat liver allografts. CXCL4, CXCR3, EGFR, JAK2, STAT3, and Collagen IV were associated with CLAD pathogenesis. We validated that CXCL4 is upstream of these informative genes in the isolated hepatic stellate cells (HSC). Blocking CXCL4 protects against CLAD by reducing liver fibrosis. Therefore, our results indicated that therapeutic approaches that neutralize CXCL4, a newly identified target of fibrosis, may represent a novel strategy for preventing and treating CLAD after liver transplantation.

Multimode optical fiber

DOEpatents

Bigot-Astruc, Marianne; Molin, Denis; Sillard, Pierre

2014-11-04

A depressed graded-index multimode optical fiber includes a central core, an inner depressed cladding, a depressed trench, an outer depressed cladding, and an outer cladding. The central core has an alpha-index profile. The depressed claddings limit the impact of leaky modes on optical-fiber performance characteristics (e.g., bandwidth, core size, and/or numerical aperture).

Femtosecond-laser inscribed double-cladding waveguides in Nd:YAG crystal: a promising prototype for integrated lasers.

PubMed

Liu, Hongliang; Chen, Feng; Vázquez de Aldana, Javier R; Jaque, D

2013-09-01

We report on the design and implementation of a prototype of optical waveguides fabricated in Nd:YAG crystals by using femtosecond-laser irradiation. In this prototype, two concentric tubular structures with nearly circular cross sections of different diameters have been inscribed in the Nd:YAG crystals, generating double-cladding waveguides. Under 808 nm optical pumping, waveguide lasers have been realized in the double-cladding structures. Compared with single-cladding waveguides, the concentric tubular structures, benefiting from the large pump area of the outermost cladding, possess both superior laser performance and nearly single-mode beam profile in the inner cladding. Double-cladding waveguides of the same size were fabricated and coated by a thin optical film, and a maximum output power of 384 mW and a slope efficiency of 46.1% were obtained. Since the large diameters of the outer claddings are comparable with those of the optical fibers, this prototype paves a way to construct an integrated single-mode laser system with a direct fiber-waveguide configuration.

General corrosion properties of modified PNC1520 austenitic stainless steel in supercritical water as a fuel cladding candidate material for supercritical water reactor

NASA Astrophysics Data System (ADS)

Nakazono, Y.; Iwai, T.; Abe, H.

2010-03-01

The Super-Critical Water-cooled Reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. There are some advantages including the use of a single phase coolant with high enthalpy but there are numerous potential problems, particularly with materials. As the operating temperature of supercritical water reactor will be between 280°C and 620°C with a pressure of 25MPa, the selection of materials is difficult and important. Austenitic stainless steels were selected for possible use in supercritical water systems because of their corrosion resistance and radiation resistance. The PNC1520 austenitic stainless steel developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. The corrosion data of PNC1520 in supercritical water (SCW) is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in supercritical water. The supercritical water corrosion test was performed for the standard PNC1520 (1520S) and the Ti-additional type of PNC1520 (1520Ti) by using a supercritical water autoclave. Corrosion tests on the austenitic 1520S and 1520Ti steels in supercritical water were performed at 400, 500 and 600°C with exposures up to 1000h. The amount of weight gain, weight loss and weight of scale were evaluated after the corrosion test in supercritical water for both austenitic steels. After 1000h corrosion test performed, the weight gains of both austenitic stainless steels were less than 2 g/m2 at 400°C and 500°C . But both weight gain and weight loss of 1520Ti were larger than those of 1520S at 600°C . By increasing the temperature to 600°C, the surface of 1520Ti was covered with magnetite formed in supercritical water and dissolution of the steel alloying elements has been observed. In view of corrosion, 1520S may have larger possibility than 1520Ti to adopt a supercritical water reactor core fuel cladding.

Microstructure and properties of Fe-based composite coating by laser cladding Fe-Ti-V-Cr-C-CeO2 powder

NASA Astrophysics Data System (ADS)

Zhang, Hui; Zou, Yong; Zou, Zengda; Wu, Dongting

2015-01-01

In situ TiC-VC reinforced Fe-based cladding layer was obtained on low carbon steel surface by laser cladding with Fe-Ti-V-Cr-C-CeO2 alloy powder. The microstructure, phases and properties of the cladding layer were investigated by X-ray diffractometry (XRD), scanning electron microscopy (SEM), energy dispersive spectrometry (EDS), transmission electron microscopy (TEM), potentio-dynamic polarization and electro-chemical impedance spectroscopy (EIS). Results showed Fe-Ti-V-Cr-C-CeO2 alloy powder formed a good cladding layer without defects such as cracks and pores. The phases of the cladding layer were α-Fe, γ-Fe, TiC, VC and TiVC2. The microstructures of the cladding layer matrix were lath martensite and retained austenite. The carbides were polygonal blocks with a size of 0.5-2 μm and distributed uniformly in the cladding layer. High resolution transmission electron microscopy showed the carbide was a complex matter composed of nano TiC, VC and TiVC2. The cladding layer with a hardness of 1030 HV0.2 possessed good wear and corrosion resistance, which was about 16.85 and 9.06 times than that of the substrate respectively.

Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

NASA Astrophysics Data System (ADS)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

2016-10-01

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

Microstructure and mechanical properties of hot wire laser clad layers for repairing precipitation hardening martensitic stainless steel

NASA Astrophysics Data System (ADS)

Wen, Peng; Cai, Zhipeng; Feng, Zhenhua; Wang, Gang

2015-12-01

Precipitation hardening martensitic stainless steel (PH-MSS) is widely used as load-bearing parts because of its excellent overall properties. It is economical and flexible to repair the failure parts instead of changing new ones. However, it is difficult to keep properties of repaired part as good as those of the substrate. With preheating wire by resistance heat, hot wire laser cladding owns both merits of low heat input and high deposition efficiency, thus is regarded as an advantaged repairing technology for damaged parts of high value. Multi-pass layers were cladded on the surface of FV520B by hot wire laser cladding. The microstructure and mechanical properties were compared and analyzed for the substrate and the clad layer. For the as-cladded layer, microstructure was found non-uniform and divided into quenched and tempered regions. Tensile strength was almost equivalent to that of the substrate, while ductility and impact toughness deteriorated much. With using laser scanning layer by layer during laser cladding, microstructure of the clad layers was tempered to fine martensite uniformly. The ductility and toughness of the clad layer were improved to be equivalent to those of the substrate, while the tensile strength was a little lower than that of the substrate. By adding TiC nanoparticles as well as laser scanning, the precipitation strengthening effect was improved and the structure was refined in the clad layer. The strength, ductility and toughness were all improved further. Finally, high quality clad layers were obtained with equivalent or even superior mechanical properties to the substrate, offering a valuable technique to repair PH-MSS.

Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

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

Unal, Cetin; Galloway, Jack D.

2014-09-12

In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermalmore » swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.« less

Revitalization of Lightweight Cladding of Buildings and Its Impact on Environment

NASA Astrophysics Data System (ADS)

Liška, Pavel; Nečasová, Barbora; Kovářová, Barbora; Novotný, Michal

2017-12-01

The presented study reveals that the revitalization of lightweight claddings installed before 1990 can have a positive impact on the environment and on the reduction of greenhouse gases in particular. The main focus is placed on the revitalization of a structural system known as OD-001, commonly called the ‘Boleticky panel’ system, which was frequently utilised all around the Czech Republic in the period before 1990. Only revitalization methods utilizing contemporary structural designs and current materials were verified during this study. The ‘Boleticky panel’ system was the type of façade cladding most frequently installed on administrative buildings in what was then Czechoslovakia. It is a panel system where load-bearing structure of the panel itself consists of closed profiles that are suspended from the building’s load-bearing structure. This type of system saw a great deal of use for more than 20 years. From today’s point of view, its thermal and technical properties are completely unsatisfactory and the gradual structural degradation of such systems, with a direct impact on their mechanical resistance, has been monitored over the last few years. However, these defects can be completely eliminated by the selection of a suitable type of revitalization. Cladding revitalization can be divided into three main categories. Each category represents a different level of impact on the structure of the above described cladding system. The first category only involves the replacement of windows, while the second consists in the replacement both of the windows and the existing panel sections. The third category of revitalization entails the complete removal of the existing cladding system and its replacement with a new one. The Life Cycle Assessment method (LCA) was used for environmental impact assessment. The aims and intentions of this method are not to search for the most economical or technically perfect product, service or technology, but to find the most environmentally friendly product with properties that can be guaranteed to last throughout its whole service life. The obtained results showed that revitalization has a positive impact on the environment. It can significantly reduce the consumption of energy that is used to heat the building in the winter, and thus reduces greenhouse gas emissions. On the other hand, it will cause a slight increase in the demand for cooling energy in the summer, which is mainly due to the reduction of the air permeability of the structure, making it more difficult to cool the interior of the building down, e.g. during the night, causing inhabitants to make greater use of air-conditioning. However, the revitalization itself, even if this term is taken to include the installation of the new cladding system, its maintenance and its future demolition, has a negligible impact on the environment compared with the old system. Therefore, based on the evaluated data the authors of the presented paper can highly recommend and encourage the revitalization of OD-001 lightweight cladding systems.

Laser materials processing of complex components: from reverse engineering via automated beam path generation to short process development cycles

NASA Astrophysics Data System (ADS)

Görgl, Richard; Brandstätter, Elmar

2017-01-01

The article presents an overview of what is possible nowadays in the field of laser materials processing. The state of the art in the complete process chain is shown, starting with the generation of a specific components CAD data and continuing with the automated motion path generation for the laser head carried by a CNC or robot system. Application examples from laser cladding and laser-based additive manufacturing are given.

Study of a Waveguide Antenna Implemented in Laminated Material

DTIC Science & Technology

2002-12-01

CRC had some experience with fabricating microwave circuits with laminated copper clad dielectric. There was bonding material (CLTE-P) that could be...PEC) and copper walls in Table 4. For all of the simulations, the physical parameters (width= 5.13 mm, height=3.40 mm, length=24.27 mm) and the... copper waveguide. As shown in Chapter 2, it is also possible to determine the losses from the conductor and dielectric (Equations 2.6 - 2.8) of a RWG

Wear Test Results of Candidate Materials for the OK-542 Towed Array Handling Machine Level Winder

DTIC Science & Technology

1994-12-29

Stainless Steel, Inconel 625 , Nickel-Aluminum-Bronze, and Titanium. The specialty materials: Inconel 625 , Monel, Stainless and Stellite, were clad-welded...metals on a base of 1040 Carbon Steel. Finally, an economic carbide coating was deposited on a 316 Stainless Steel and Inconel 625 sample. Within a...damage in the shortest period of time. The Inconel 625 bar stock that was tested performed the best. It sustained the least amount of damage for one

Microstructured Optical Fiber for X-ray Detection

NASA Technical Reports Server (NTRS)

DeHaven, Stanton L.

2009-01-01

A novel scintillating optical fiber is presented using a composite micro-structured quartz optical fiber. Scintillating materials are introduced into the multiple inclusions of the fiber. This creates a composite optical fiber having quartz as a cladding with an organic scintillating material core. X-ray detection using these fibers is compared to a collimated cadmium telluride (CdTe) detector over an energy range from 10 to 40 keV. Results show a good correlation between the fiber count rate trend and that of the CdTe detector.

Phosphate-core silica-clad Er/Yb-doped optical fiber and cladding pumped laser.

PubMed

Egorova, O N; Semjonov, S L; Velmiskin, V V; Yatsenko, Yu P; Sverchkov, S E; Galagan, B I; Denker, B I; Dianov, E M

2014-04-07

We present a composite optical fiber with a Er/Yb co-doped phosphate-glass core in a silica glass cladding as well as cladding pumped laser. The fabrication process, optical properties, and lasing parameters are described. The slope efficiency under 980 nm cladding pumping reached 39% with respect to the absorbed pump power and 28% with respect to the coupled pump power. Due to high doping level of the phosphate core optimal length was several times shorter than that of silica core fibers.

Examination of T-111 clad uranium nitride fuel pins irradiated up to 13,000 hours at a clad temperature of 990 C

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.

1973-01-01

The examination of 27 fuel pins irradiated for up to 13,000 hours at 990 C is described. The fuel pin clad was a tantalum alloy with uranium nitride as the nuclear fuel. Two nominal fuel pin diameters were tested with a maximum burnup of 2.34 atom percent. Twenty-two fuel pins were tested for fission gas leaks; thirteen pins leaked. Clad ductility tests indicated clad embrittlement. The embrittlement is attributed to hydrogen from an n,p reaction in the fuel. Fuel swelling was burnup dependent, and the amount of fission gas release was low, generally less than 0.5 percent. No incompatibilities between fuel, liner, and clad were in evidence.

A modified hexagonal photonic crystal fiber for terahertz applications

NASA Astrophysics Data System (ADS)

Islam, Md. Saiful; Sultana, Jakeya; Faisal, Mohammad; Islam, Mohammad Rakibul; Dinovitser, Alex; Ng, Brian W.-H.; Abbott, Derek

2018-05-01

We present a Zeonex based highly birefringent and dispersion flattened porous core photonic crystal fiber (PC-PCF) for polarization preserving applications in the terahertz region. In order to facilitate birefringence, an array of elliptical shaped air holes surrounded by porous cladding is introduced. The porous cladding comprises circular air-holes in a modified hexagonal arrangement. The transmission characteristics of the proposed PCF are investigated using a full-vector finite element method with perfectly matched layer (PML) absorbing boundary conditions. Simulation results show a high birefringence of 0.086 and an ultra-flattened dispersion variation of ± 0.03 ps/THz/cm at optimal design parameters. Besides, a number of other important wave-guiding properties including frequency dependence of the effective material loss (EML), confinement loss, and effective area are also investigated to assess the fiber's effectiveness as a terahertz waveguide.

Development of a highly reliable composite board for printed circuitry for use in space environment

NASA Technical Reports Server (NTRS)

Bradbury, E. J.; Markle, R. A.; Dunnavant, W. R.; Stickney, P. B.

1971-01-01

Materials, processes and fabrication techniques have been investigated for the development of a high-temperature circuit-board laminate. High quality, void-free copper-clad laminates have been made using 7628/HS-1 style fiberglas reinforcements with filled polyimide matrices. The fabricating characteristics of P13N resin appear suitable for use as a filled matrix in this circuit board development. High-fired, ball-milled alumina appears to be necessary to obtain the desired effects in the circuit board system. Nickel-clad copper foil bonding surfaces appear to be another requirement for retention of good bond strengths after art work and plating sequences. The fabrication cycle for this circuit board system is very dependent on the heating profile. Very rapid heating with quick loading is recommended. A stack approach to lamination was successfully used.

Characterization of long-range plasmonic waveguides at visible to near-infrared regime

NASA Astrophysics Data System (ADS)

Huang, Sheng-Ting; Lai, Chien-Chih; Sheu, Fang-Wen; Tsai, Wan-Shao

2017-12-01

Long-range surface plasmon polariton waveguides composed with thin gold stripes embedded in SU-8 polymer cladding with various stripe widths were fabricated. Material properties of the polymer cladding layer, gold thin film, and the device structures were discussed. Optical properties based on modal propagation were characterized at visible to near-infrared wavelengths. The measured propagation losses of waveguide widths from 3 to 9 μm at 633, 785, and 1550 nm are 7.5-18.8, 6.8-12.5, and 1.9-3.9 dB/mm, respectively. Guiding mode properties such as overlap integrals between the simulated and the measured fields and the polarization extinction ratios of the waveguides with different stripe widths were investigated at the telecommunication wavelength. Good accordance between the measurement and simulation results was presented.

Expanded plug method for developing circumferential mechanical properties of tubular materials

DOEpatents

Hendrich, William Ray; McAfee, Wallace Jefferson; Luttrell, Claire Roberta

2006-11-28

A method for determining the circumferential properties of a tubular product, especially nuclear fuel cladding, utilizes compression of a polymeric plug within the tubular product to determine strain stress, yield stress and other properties. The process is especially useful in the determination of aging properties such as fuel rod embrittlement after long burn-down.

Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

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

Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.

A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

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

Heuser, Brent; Stubbins, James; Kozlowski, Tomasz

The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys.more » The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be. International fabrication options were explored in Europe and Asia, but this proved to be impractical, if not impossible. Consequently, experimental investigation of the Zr-Be binary system was dropped and exploration binary Zr-Y binary system was initiated. The motivation behind the Zr-Y system is the known thermodynamic stability of yttria over zirconia.« less

Numerical Simulation of the Emergency Condenser of the SWR-1000

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

Krepper, Eckhard; Schaffrath, Andreas; Aszodi, Attila

The SWR-1000 is a new innovative boiling water reactor (BWR) concept, which was developed by Siemens AG. This concept is characterized in particular by passive safety systems (e.g., four emergency condensers, four building condensers, eight passive pressure pulse transmitters, and six gravity-driven core-flooding lines). In the framework of the BWR Physics and Thermohydraulic Complementary Action to the European Union BWR Research and Development Cluster, emergency condenser tests were performed by Forschungszentrum Juelich at the NOKO test facility. Posttest calculations with ATHLET are presented, which aim at the determination of the removable power of the emergency condenser and its operation mode.more » The one-dimensional thermal-hydraulic code ATHLET was extended by the module KONWAR for the calculation of the heat transfer coefficient during condensation in horizontal tubes. In addition, results of conventional finite difference calculations using the code CFX-4 are presented, which investigate the natural convection during the heatup process at the secondary side of the NOKO test facility.« less

Optimization of a Boiling Water Reactor Loading Pattern Using an Improved Genetic Algorithm

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

Kobayashi, Yoko; Aiyoshi, Eitaro

2003-08-15

A search method based on genetic algorithms (GA) using deterministic operators has been developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). The search method uses an Improved GA operator, that is, crossover, mutation, and selection. The handling of the encoding technique and constraint conditions is designed so that the GA reflects the peculiar characteristics of the BWR. In addition, some strategies such as elitism and self-reproduction are effectively used to improve the search speed. LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and three-dimensional-dependent constraints have alwaysmore » necessitated the use of three-dimensional core simulators for BWRs, so that an optimization method is required for computational efficiency. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant applying the Haling technique. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

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

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

2013-11-10

A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic tomore » initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.« less

Hot Forging of a Cladded Component by Automated GMAW Process

NASA Astrophysics Data System (ADS)

Rafiq, Muhammad; Langlois, Laurent; Bigot, Régis

2011-01-01

Weld cladding is employed to improve the service life of engineering components by increasing corrosion and wear resistance and reducing the cost. The acceptable multi-bead cladding layer depends on single bead geometry. Hence, in first step, the relationship between input process parameters and the single bead geometry is studied and in second step a comprehensive study on multi bead clad layer deposition is carried out. This paper highlights an experimental study carried out to get single layer cladding deposited by automated Gas Metal Arc Welding (GMAW) process and to find the possibility of hot forming of the cladded work piece to get the final hot formed improved structure. GMAW is an arc welding process that uses an arc between a consumable electrode and the welding pool with an external shielding gas and the cladding is done by alongside deposition of weld beads. The experiments for single bead were conducted by varying the three main process parameters wire feed rate, arc voltage and welding speed while keeping other parameters like nozzle to work distance, shielding gas and its flow rate and torch angle constant. The effect of bead spacing and torch orientation on the cladding quality of single layer from the results of single bead deposition was studied. Effect of the dilution rate and nominal energy on the cladded layer hot bending quality was also performed at different temperatures.

Effects of Laser Power Level on Microstructural Properties and Phase Composition of Laser-Clad Fluorapatite/Zirconia Composite Coatings on Ti6Al4V Substrates.

PubMed

Chien, Chi-Sheng; Liu, Cheng-Wei; Kuo, Tsung-Yuan

2016-05-17

Hydroxyapatite (HA) is one of the most commonly used materials for the coating of bioceramic titanium (Ti) alloys. However, HA has poor mechanical properties and a low bonding strength. Accordingly, the present study replaces HA with a composite coating material consisting of fluorapatite (FA) and 20 wt % yttria (3 mol %) stabilized zirconia (ZrO₂, 3Y-TZP). The FA/ZrO₂ coatings are deposited on Ti6Al4V substrates using a Nd:YAG laser cladding system with laser powers and travel speeds of 400 W/200 mm/min, 800 W/400 mm/min, and 1200 W/600 mm/min, respectively. The experimental results show that a significant inter-diffusion of the alloying elements occurs between the coating layer (CL) and the transition layer (TL). Consequently, a strong metallurgical bond is formed between them. During the cladding process, the ZrO₂ is completely decomposed, while the FA is partially decomposed. As a result, the CLs of all the specimens consist mainly of FA, Ca₄(PO₄)₂O (TTCP), CaF₂, CaZrO₃, CaTiO₃ and monoclinic phase ZrO₂ (m-ZrO₂), together with a small amount of θ-Al₂O₃. As the laser power is increased, CaO, CaCO₃ and trace amounts of tetragonal phase ZrO₂ (t-ZrO₂) also appear. As the laser power increases from 400 to 800 W, the CL hardness also increases as a result of microstructural refinement and densification. However, at the highest laser power of 1200 W, the CL hardness reduces significantly due to the formation of large amounts of relatively soft CaO and CaCO₃ phase.

Technique development for modulus, microcracking, hermeticity, and coating evaluation capability characterization of SiC/SiC tubes

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

Hu, Xunxiang; Ang, Caen K.; Singh, Gyanender P.

Driven by the need to enlarge the safety margins of nuclear fission reactors in accident scenarios, research and development of accident-tolerant fuel has become an important topic in the nuclear engineering and materials community. A continuous-fiber SiC/SiC composite is under consideration as a replacement for traditional zirconium alloy cladding owing to its high-temperature stability, chemical inertness, and exceptional irradiation resistance. An important task is the development of characterization techniques for SiC/SiC cladding, since traditional work using rectangular bars or disks cannot directly provide useful information on the properties of SiC/SiC composite tubes for fuel cladding applications. At Oak Ridge Nationalmore » Laboratory, experimental capabilities are under development to characterize the modulus, microcracking, and hermeticity of as-fabricated, as-irradiated SiC/SiC composite tubes. Resonant ultrasound spectroscopy has been validated as a promising technique to evaluate the elastic properties of SiC/SiC composite tubes and microcracking within the material. A similar technique, impulse excitation, is efficient in determining the basic mechanical properties of SiC bars prepared by chemical vapor deposition; it also has potential for application in studying the mechanical properties of SiC/SiC composite tubes. Complete evaluation of the quality of the developed coatings, a major mitigation strategy against gas permeation and hydrothermal corrosion, requires the deployment of various experimental techniques, such as scratch indentation, tensile pulling-off tests, and scanning electron microscopy. In addition, a comprehensive permeation test station is being established to assess the hermeticity of SiC/SiC composite tubes and to determine the H/D/He permeability of SiC/SiC composites. This report summarizes the current status of the development of these experimental capabilities.« less

The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions

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

Cinbiz, Mahmut N; Brown, Nicholas R; Terrani, Kurt A

2017-01-01

This study investigates the failure mechanisms of advanced nuclear fuel cladding of FeCrAl at high-strain rates, similar to design basis reactivity initiated accidents (RIA). During RIA, the nuclear fuel cladding was subjected to the plane-strain to equibiaxial tension strain states. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses as occurring RIA. The mechanical response of the advanced claddings was compared to that of hydrided zirconium-based nuclear fuel cladding alloy. The hoop strain evolution during pressure pulses were collected in situ; the permanent diametral strains of bothmore » accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture.« less

Cyclic furnace oxidation of clad WI-52 systems at 1040 C and 1090 C

NASA Technical Reports Server (NTRS)

Gedwill, M. A.

1972-01-01

Cyclic furnace oxidation studies were conducted on the cobalt alloy WI-52 clad with Ni-30Cr, Fe-25Cr-4A1, and Ni-20Cr-4A1 foils (0.051 to 0.254 mm thick). Tests as long as 400 hours using 1- and 20-hour cycles showed that the Ni-Cr- and Fe-Cr-A1 claddings were about equally protective at both temperatures. The protective ability of these alloys was influenced by exposure temperature and cladding thickness. At both temperatures, they protected WI-52 about as well as, or better than, a widely used commercial aluminide coating. The Ni-Cr-Al claddings did not protect WI-52 nearly as well. Interdiffusion generally influenced the oxidation behavior of all clad WI-52 systems.

Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

NASA Astrophysics Data System (ADS)

Artnak, Edward Joseph, III

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

Protective claddings for high strength chromium alloys

NASA Technical Reports Server (NTRS)

Collins, J. F.

1971-01-01

The application of a Cr-Y-Hf-Th alloy as a protective cladding for a high strength chromium alloy was investigated for its effectiveness in inhibiting nitrogen embrittlement of a core alloy. Cladding was accomplished by a combination of hot gas pressure bonding and roll cladding techniques. Based on bend DBTT, the cladding alloy was effective in inhibiting nitrogen embrittlement of the chromium core alloy for up to 720 ks (200hours) in air at 1422 K (2100 F). A significant increase in the bend DBTT occurred with longer time exposures at 1422 K or short time exposures at 1589 K (2400 F).

Orientation-Dependent Displacement Sensor Using an Inner Cladding Fiber Bragg Grating

PubMed Central

Yang, Tingting; Qiao, Xueguang; Rong, Qiangzhou; Bao, Weijia

2016-01-01

An orientation-dependent displacement sensor based on grating inscription over a fiber core and inner cladding has been demonstrated. The device comprises a short piece of multi-cladding fiber sandwiched between two standard single-mode fibers (SMFs). The grating structure is fabricated by a femtosecond laser side-illumination technique. Two well-defined resonances are achieved by the downstream both core and cladding fiber Bragg gratings (FBGs). The cladding resonance presents fiber bending dependence, together with a strong orientation dependence because of asymmetrical distribution of the “cladding” FBG along the fiber cross-section. PMID:27626427

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Experimental Investigation

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

Wirth, Brian; Morgan, Dane; Kaoumi, Djamel

2013-12-01

The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, theirmore » evolution is highly interlinked. Radiationinduced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not yet possible as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution under irradiation. This project will focus on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ experiments that can provide validation and benchmarking to the computer codes. The broad scientific and technical objectives of this proposal are to evaluate the microstructure and microchemical evolution in advanced ferritic/martensitic and oxide dispersion strengthened (ODS) alloys for cladding and duct reactor materials under long-term and elevated temperature irradiation, leading to improved ability to model structural materials performance and lifetime. Specifically, we propose four research thrusts, namely Thrust 1: Identify the formation mechanism and evolution for dislocation loops with Burgers vector of a<100> and determine whether the defect microstructure (predominately dislocation loop/dislocation density) saturates at high dose. Thrust 2: Identify whether a threshold irradiation temperature or dose exists for the nucleation of growing voids that mark the beginning of irradiation-induced swelling, and begin to probe the limits of thermal stability of the tempered Martensitic structure under irradiation. Thrust 3: Evaluate the stability of nanometer sized Y- Ti-O based oxide dispersion strengthened (ODS) particles at high fluence/temperature. Thrust 4: Evaluate the extent to which precipitates form and/or dissolve as a function of irradiation temperature and dose, and how these changes are driven by radiation induced segregation and microchemical evolutions and determined by the initial microstructure.« less

Microstructure and Corrosion Behavior of Laser Synthesized Cobalt Based Powder on Ti-6Al-4V

NASA Astrophysics Data System (ADS)

Adesina, O. S.; Popoola, A. P. I.; Pityana, S. L.; Oloruntoba, D. T.

2018-05-01

The corrosion behavior of titanium alloys when used for various dynamic offshore components has been a major concern of titanium drilling risers in deepwater energy extraction. A way of achieving specified requirement is the development of coatings suitable to protect the base material against corrosion. In this work, laser cladding technique which is known as a leading edge due to its distinctive properties and outcomes was used in synthesizing Co-based powder on titanium alloy. The processing parameters used were laser power of 900W; scan speed of 0.6 to 1.2 m/min; powderfeedrate1.0g/min;beamspotsize3mm;gasflowrate1.2L/min.The effects of cobalt addition and laser parameters on corrosion behavior of laser clad Ti6AL4V coating in 0.5M sulfuric medium were investigated using linear potentiodynamic polarization. The changes in microstructure and corrosion behavior were analyzed using scanning electron microscopy (SEM) while the X –ray diffraction (XRD) indicates the intermetallics in the coatings. Results showed that the coatings displayed good metallurgical bonding with dendritic formations between the coatings and the substrate. The anodic current density increased with lower scan speed. However, the corrosion current densities of laser-clad samples were lower than Ti6Al4V alloy.

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development

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

George W. Griffith

2011-10-01

A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the fibers and the matrix allows formore » ductile behavior. The SiC CMC has relatively high strength at high reactor accident temperatures when compared to metallic cladding. SiC also has a very low chemical reactivity and doesn't react exothermically with the reactor cooling water. The radiation behavior of SiC has also been studied extensively as structural fusion system components. The SiC CMC technology is in the early stages of development and will need to mature before confidence in the developed designs can created. The advanced SiC CMC materials do offer the potential for greatly improved safety because of their high temperature strength, chemical stability and reduced hydrogen generation.« less

Experimental Study of Two Phase Flow Behavior Past BWR Spacer Grids

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

Ratnayake, Ruwan K.; Hochreiter, L.E.; Ivanov, K.N.

2002-07-01

Performance of best estimate codes used in the nuclear industry can be significantly improved by reducing the empiricism embedded in their constitutive models. Spacer grids have been found to have an important impact on the maximum allowable Critical Heat Flux within the fuel assembly of a nuclear reactor core. Therefore, incorporation of suitable spacer grids models can improve the critical heat flux prediction capability of best estimate codes. Realistic modeling of entrainment behavior of spacer grids requires understanding the different mechanisms that are involved. Since visual information pertaining to the entrainment behavior of spacer grids cannot possibly be obtained frommore » operating nuclear reactors, experiments have to be designed and conducted for this specific purpose. Most of the spacer grid experiments available in literature have been designed in view of obtaining quantitative data for the purpose of developing or modifying empirical formulations for heat transfer, critical heat flux or pressure drop. Very few experiments have been designed to provide fundamental information which can be used to understand spacer grid effects and phenomena involved in two phase flow. Air-water experiments were conducted to obtain visual information on the two-phase flow behavior both upstream and downstream of Boiling Water Reactor (BWR) spacer grids. The test section was designed and constructed using prototypic dimensions such as the channel cross-section, rod diameter and other spacer grid configurations of a typical BWR fuel assembly. The test section models the flow behavior in two adjacent sub channels in the BWR core. A portion of a prototypic BWR spacer grid accounting for two adjacent channels was used with industrial mild steel rods for the purpose of representing the channel internals. Symmetry was preserved in this practice, so that the channel walls could effectively be considered as the channel boundaries. Thin films were established on the rod surfaces by injecting water through a set of perforations at the bottom ends of the rods, ensuring that the flow upstream of the bottom-most spacer grid is predominantly annular. The flow conditions were regulated such that they represent typical BWR operating conditions. Photographs taken during experiments show that the film entrainment increases significantly at the spacer grids, since the points of contact between the rods and the grids result in a peeling off of large portions of the liquid film from the rod surfaces. Decreasing the water flow resulted in eventual drying out, beginning at positions immediately upstream of the spacer grids. (authors)« less

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

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

Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M

2015-01-01

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k eff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of latticemore » design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.« less