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Evaluation of liquid metal protection of a limiter/divertor in fusion reactors

Description: The liquid metal protection concept is proposed mainly to prolong the lifetime of a divertor or a limiter in a fusion reactor. This attractive idea for protection requires studying a wide range of problems associated with the use of liquid-metals in fusion reactors. In this work the protection by liquid-metals has concentrated on predictions of the loss rate of the film to the plasma, the operating surface temperatures required for the film, and the potential tritium inventory requirement. The effect of plasma disruptions on the liquid metal film is also evaluated. Other problems such as liquid metal compatibility with structural materials, magnetic field effects, and the effect of liquid metal contamination on plasma performance are discussed. Three candidate liquid-metals are evaluated, i.e., lithium, gallium, and tin. A wide range of reactor operating conditions valid for both near term machines (INTOR and ITER) and for the next generation commercial reactors (TPSS) are considered. This study has indicated that the evaporation rate for candidate liquid metals can be kept below the sputtering range for reasonable operating temperatures and plasma edge conditions. At higher temperatures, evaporation dominates the losses. Impurity transport calculations indicate that impurities from the plate should not reach the main plasma. One or two millimeters of liquid films can protect the structure from severe plasma disruptions. Depending on the design of the liquid metal protection system, the tritium inventory in the liquid film is predicted to be on the order of a few grams. 16 refs., 5 figs.
Date: January 1, 1988
Creator: Hassanein, A.M. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Corrosion of ferrous alloys in eutectic lead-lithium environments

Description: Corrosion data have been obtained on austenitic prime candidate alloy (PCA) and Type 316 stainless steel and ferritic HT-9 and Fe-9Cr-1Mo steels in a flowing Pb-17 at. % Li environment at 727 and 700 K (454 and 427/sup 0/C). The results indicate that the dissolution rates for both austenitic and ferritic steels in Pb-17Li are an order of magnitude greater than in flowing lithium. The influence of time, temperature, and alloy composition on the corrosion behavior in Pb-17Li is similar to that in lithium. The weight losses for the austenitic steels are an order of magnitude greater than for the ferritic steels. The rate of weight loss for the ferritic steels is constant, whereas the dissolution rates for the austenitic steels decrease with time. After exposure to Pb-17Li, the austenitic steels develop a very weak and porous ferrite layer which easily spalls from the specimen surface.
Date: September 1, 1983
Creator: Chopra, O.K. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Influence of a flowing-lithium environment on the fatigue and tensile properties of Type 316 stainless steel

Description: Low-cycle fatigue and tensile data have been obtained on Type 316 stainless steel in a flowing lithium environment of controlled purity. The results show that the fatigue life of the steel in flowing lithium at 755 K is greater than in air. Preexposure of the material to lithium reduces fatigue life. The reduction in fatigue life may be attributed to the formation of a weak ferrite layer after lithium exposure. Tensile data for cold-worked Type 316 stainless steel indicate that at temperatures between 476 and 755 K, a flowing lithium environment has little or no effect on the tensile properties of the steel.
Date: September 1, 1983
Creator: Chopra, O.K. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Generation of covariance data among values from a single set of experiments

Description: Modern nuclear data evaluation methods demand detailed uncertainty information for all input results to be considered. It can be shown from basic statistical principles that provision of a covariance matrix for a set of data provides the necessary information for its proper consideration in the context of other included experimental data and/or a priori representations of the physical parameters in question. This paper examines how an experimenter should go about preparing the covariance matrix for any single experimental data set he intends to report. The process involves detailed examination of the experimental procedures, identification of all error sources (both random and systematic); and consideration of any internal discrepancies. Some specific examples are given to illustrate the methods and principles involved.
Date: January 1, 1992
Creator: Smith, D.L.
Partner: UNT Libraries Government Documents Department

Low-cycle fatigue behavior of HT-9 alloy in a flowing-lithium environment

Description: Low-cycle fatigue data have been obtained on normalized/tempered or lithium-preexposed HT-9 alloy at 755 K in flowing lithium of controlled purity. The results show that the fatigue life of this material decreases with an increase in nitrogen content in lithium. A reduction in strain rate also decreases the fatigue life in high-nitrogen lithium. However, in the range from approx. 4 x 10/sup -4/ to 4 x 10/sup -2/ s/sup -1/, the strain rate has no effect on fatigue life in lithium containing <200 wppM nitrogen. The fatigue life of the HT-9 alloy in low-nitrogen lithium is significantly greater than the fatigue life of Fe-9Cr-1Mo steel or Type 403 martensitic steel in air. Furthermore, a 4.0-Ms preexposure to low-nitrogen lithium has no influence on fatigue life. The reduction in fatigue life in high-nitrogen lithium is attributed to internal corrosive attack of the material. The specimens tested in high-nitrogen lithium show internal corrosion along grain and martensitic lathe boundaries and intergranular fracture. This behavior is not observed in specimens tested in low-nitrogen lithium. Results for a constant-load corrosion test in flowing lithium are also presented.
Date: June 1, 1983
Creator: Chopra, O.K. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Neutron inelastic scattering studies for lead-204. [Threshold energy to 10 MeV, transitions, cross sections, angular distributions]

Description: A 9.57-g sample of lead metal, enriched to 99.7 percent /sup 204/Pb, was used in an investigation of neutron inelastic scattering from this rare isotope at the Argonne National Laboratory Fast-Neutron Generator Facility. Neutron excitation of the 66.9-m isomeric state at 2.186 MeV in /sup 204/Pb was measured from near threshold to approximately 10 MeV using activation techniques. Cross sections and a value for the isomeric half life were derived from these data. Time-of-flight techniques were employed to measure spectra of promptly-emitted gamma rays from the /sup 204/Pb(n;n',..gamma..)/sup 204/Pb reaction at neutron energies less than or equal to 3 MeV. Cross sections and angular distributions were derived from these data for several of the stronger transitions. Other available fast-neutron data for this isotope are reviewed briefly.
Date: December 1, 1977
Creator: Smith, D.L. & Meadows, J.W.
Partner: UNT Libraries Government Documents Department

Fast neutron radiative capture cross section of /sup 232/Th. [30 keV to 2. 5 MeV]

Description: The /sup 232/Th(n,..gamma..) cross section was measured between 30 keV and 2.5 MeV. A large liquid scintillator was used to measure the shape of the cross section relative to Au(n,..gamma..) between 58 keV and 850 keV. The activation technique was used in measurements at 30 keV relative to Au(n,..gamma..) and above 240 keV relative to /sup 235/U(n,f). The activation results were utilized to normalize the shape data. The results agree well with recent experimental data by Lindner et al. and are substantially lower than the evaluated data file ENDF/B-IV in the several-hundred-keV range.
Date: March 1, 1978
Creator: Poenitz, W.P. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Elastic and inelastic surface effects on ion penetration and the resulting sputtering and backscattering

Description: The computer code ITMC (Ion Transport in Materials and Compounds) has been developed to study in detail the transport of charged particles in solid materials and surface related phenomena such as sputtered atoms and backscattered ions. The code is based on Monte Carlo methods to follow the path and the damage produced by the charged particles in three dimension as they slow down in target materials. Single-element targets as well as alloys with possible different surface and bulk compositions or with layered structures of different materials can be used. Various models developed to calculate the inelastic energy losses with target electrons can be used in the code. Most known interatomic potentials can also be used to calculate the elastic energy losses. The major advantages of the code are its ability and flexibility to use and compare various models of elastic and inelastic energy losses in any target with different compounds and different surface and bulk composition.
Date: January 1, 1985
Creator: Hassanein, A.M. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Stress-corrosion cracking susceptibility of V-15Cr-5Ti in pressurized water at 288/sup 0/C

Description: The stress-corrosion cracking susceptibility of V-15Cr-5Ti in pressurized water at 288/sup 0/C has been evaluated by means of constant extension rate tensile (CERT) tests in a refreshed autoclave system. The test environments included high-purity water as well as water containing SO/sub 4//sup 2 -/ and NO/sub 3//sup -/ impurities at a concentration of 10 wppM. Strain rates from 1 x 10/sup -6/ to 5 x 10/sup -8/ s/sup -1/ were employed, and dissolved oxygen levels ranged from <0.005 to 7.9 wppM. Test times were from 3.2 to 619 h. No stress corrosion cracking was observed under any of the test conditions. These results were analyzed using measured electrochemical potentials, available Pourbaix diagram information, and the observed oxidation behavior. 7 refs., 5 figs., 1 tab.
Date: July 1, 1987
Creator: Diercks, D.R. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Influence of the deuteron energy on the testing volume of IFMIF and its impact on other parameters

Description: The influence of the energy of the deuteron beam on irradiation parameters of IFMIF is analyzed. The main purpose of this paper is to identify possible positive and negative impacts on irradiation parameters that an increase in the deuteron energy of the beam can cause. Several parameters of the facility, such as neutron generation rate, number of neutrons with energy above 20 MeV at the source and in the test assembly, volume with dpa rate above a threshold value, gas production, and gradient of the atomic displacement rate, are analyzed and conclusions are drawn based on the calculated values. It is shown that an increase in the deuteron energy to 40 MeV does not produce a significant negative impact for the elements analyzed, but instead is beneficial in producing nuclear responses more similar to a fusion environment than the lower deuteron energies.
Date: September 1, 1995
Creator: Gomes, I.C. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Decontamination and dismantlement of the old hot laundry, CFA-669. Final report

Description: This final report describes the decontamination and dismantlement (D&D) of the old hot laundry, located at the Idaho National Engineering Laboratory (INEL) Central Facilities Area (CFA). The report describes the site before and after D&D, processes used, cost and duration, and waste volume generated. In addition, lessons learned are presented. Pre-D&D characterization indicated gross alpha concentrations in the building ranged from 6 to 310 pCi/g and gross beta measurements from 6 to 15,000 pCi/g. Gamma spectrum analysis identified cobalt-60, cesium-137, antimony-125, europium-152, europium-154, and niobium-94.
Date: January 1, 1995
Creator: Smith, D.L. & Perry, E.F.
Partner: UNT Libraries Government Documents Department

Nuclear data needs for non-intrusive inspection.

Description: Various nuclear-based techniques are being explored for use in non-intrusive inspection. Their development is motivated by the need to prevent the proliferation of nuclear weapons, to thwart trafficking in illicit narcotics, to stop the transport of explosives by terrorist organizations, to characterize nuclear waste, and to deal with various other societal concerns. Non-intrusive methods are sought in order to optimize inspection speed, to minimize damage to packages and containers, to satisfy environmental, health and safety requirements, to adhere to legal requirements, and to avoid inconveniencing the innocent. These inspection techniques can be grouped into two major categories: active and passive. They almost always require the use of highly penetrating radiation and therefore are generally limited to neutrons and gamma rays. Although x-rays are widely employed for these purposes, their use does not constitute nuclear technology and therefore is not discussed here. This paper examines briefly the basic concepts associated with nuclear inspection and investigates the related nuclear data needs. These needs are illustrated by considering four of the methods currently being developed and tested.
Date: November 29, 2000
Creator: Smith, D. L. & Michlich, B. J.
Partner: UNT Libraries Government Documents Department

Multiplier, moderator, and reflector materials for lithium-vanadium fusion blankets.

Description: The self-cooled lithium-vanadium fusion blanket concept has several attractive operational and environmental features. In this concept, liquid lithium works as the tritium breeder and coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because of its superior performance relative to other alloys for this application. However, this concept has poor attenuation characteristics and energy multiplication for the DT neutrons. An advanced self-cooled lithium-vanadium fusion blanket concept has been developed to eliminate these drawbacks while maintaining all the attractive features of the conventional concept. An electrical insulator coating for the coolant channels, spectral shifter (multiplier, and moderator) and reflector were utilized in the blanket design to enhance the blanket performance. In addition, the blanket was designed to have the capability to operate at high loading conditions of 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading. This paper assesses the spectral shifter and the reflector materials and it defines the technological requirements of this advanced blanket concept.
Date: October 7, 1999
Creator: Gohar, Y. & Smith, D. L.
Partner: UNT Libraries Government Documents Department

A compilation of information on the {sup 31}P(p,{gamma}){sup 32}S reaction and properties of excited levels in {sup 32}S

Description: This report documents a survey of the literature, and provides a compilation of data contained therein, for the {sup 31}P(p,{gamma}){sup 32}S reaction. Attention here is paid mainly to resonance states in the compound-nuclear system {sup 32}S formed by {sup 31}P + p, with emphasis on radiative capture, i.e., gamma-ray decay channels ({sup 32}Si + {gamma}) which populate specific levels in {sup 32}S. The energy region near the proton separation energy for {sup 32}S is especially important in this context for applications in nuclear astrophysics. Properties of the excited states in {sup 32}S are also considered. Summaries of all the located references with significant content are provided and numerical data contained in them are compiled in EXFOR format where applicable.
Date: May 31, 2000
Creator: Smith, D. L. & Daly, J. T.
Partner: UNT Libraries Government Documents Department

Nuclear Data and Measurement Series - a method to construct covariance files in ENDF/B format for criticality safety applications.

Description: Argonne National Laboratory is providing support for a criticality safety analysis project that is being performed at Oak Ridge National Laboratory. The ANL role is to provide the covariance information needed by ORNL for this project. The ENDF/B-V evaluation is being used for this particular criticality analysis. In this evaluation, covariance information for several isotopes or elements of interest to this analysis is either not given or needs to be reconsidered. For some required materials, covariance information does not exist in ENDF/B-V: {sup 233}U, {sup 236}U, Zr, Mg, Gd, and Hf. For others, existing covariance information may need to be re-examined in light of the newer ENDF/B-V evaluation and recent experimental data. In this category are the following materials: {sup 235}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, Fe, H, C, N, O, Al, Si, and B. A reasonable estimation of the fractional errors for various evaluated neutron cross sections from ENDF/B-V can be based on the comparisons between the major more recent evaluations including ENDF/B-VI, JENDL3.2, BROND2.2, and JEF2.2, as well as a careful examination of experimental data. A reasonable method to construct correlation matrices is proposed here. Coupling both of these considerations suggests a method to construct covariances files in ENDF/B format that can be used to express uncertainties for specific ENDF/B-V cross sections.
Date: July 30, 1999
Creator: Naberejnev, D.G. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Fast-neutron-spectrum measurements for the thick-target /sup 9/Be(d,n)/sup 10/B reaction at E/sub d/ = 7 MeV

Description: Spectra of neutrons with energies greater than or equal to 800 keV which are emitted from a metallic beryllium target that is thick enough to completely stop 7-MeV incident deuterons are measured using organic scintillators and the pulse-beam time-of-flight method. Data are acquired for twenty different emission angles in the laboratory over the range 0 to 155 degrees. The resulting information on the energy-angle detail of neutron emission is then employed in calculations which are performed in order to examine certain effects of the anisotropic neutron production on typical measurements of integral fast-neutron reaction cross sections. 28 refs.
Date: April 1, 1985
Creator: Smith, D.L.; Meadows, J.W. & Guenther, P.T.
Partner: UNT Libraries Government Documents Department

Vanadium-base alloys for fusion reactor applications

Description: Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined.
Date: October 1, 1984
Creator: Smith, D.L.; Loomis, B.A. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Evaluated nuclear-data file for niobium

Description: A comprehensive evaluated nuclear-data file for elemental niobium is provided in the ENDF/B format. This file, extending over the energy range 10/sup -11/-20 MeV, is suitable for comprehensive neutronic calculations, particulary those dealing with fusion-energy systems. It also provides dosimetry information. Attention is given to the internal consistancy of the file, energy balance, and the quantitative specification of uncertainties. Comparisons are made with experimental data and previous evaluated files. The results of integral tests are described and remaining outstanding problem areas are cited. 107 refs.
Date: March 1, 1985
Creator: Smith, A.B.; Smith, D.L. & Howerton, R.J.
Partner: UNT Libraries Government Documents Department

Compilation and evaluation of 14-MeV neutron-activation cross sections for nuclear technology applications. Set I

Description: Available 14-MeV experimental neutron activation cross sections are compiled and evaluated for the following reactions of interest for nuclear-energy technology applications: /sup 27/Al(n,p)/sup 27/Mg, Si(n,X)/sup 28/Al, Ti(n,X)/sup 46/Sc, Ti(n,X)/sup 47/Sc, Ti(n,X)/sup 48/Sc, /sup 51/V(n,p)/sup 51/Ti, /sup 51/V(n,..cap alpha..)/sup 48/Sc, Cr(n,X)/sup 52/V, /sup 55/Mn(n,..cap alpha..)/sup 52/V, /sup 55/Mn(n,2n)/sup 54/Mn, Fe(n,X)/sup 54/Mn, /sup 54/Fe(n,..cap alpha..)/sup 51/Cr, /sup 59/Co(n,p)/sup 59/Fe, /sup 59/Co(n,..cap alpha..)/sup 56/Mn, /sup 59/Co(n,2n)/sup 58/Co, /sup 65/Cu(n,p)/sup 65/Ni, Zn(n,X)/sup 64/Cu, /sup 64/Zn(n,2n)/sup 63/Zn, /sup 113/In(n,n')/sup 113m/In, /sup 115/In(n,n') /sup 115m/In. The compiled values are listed and plotted for reference without adjustments. From these collected results those values for which adequate supplementary information on nuclear constants, standards and experimental errors is provided are selected for use in reaction-by-reaction evaluations. These data are adjusted as needed to account for recent revisions in the nuclear constants and cross section standards. The adjusted results are subsequently transformed to equivalent cross sections at 14.7 MeV for the evaluation process. The evaluations are performed utilizing a least-squares method which considers correlations between the experimental data. 440 refs., 41 figs., 46 tabs.
Date: April 1, 1985
Creator: Evain, B.P.; Smith, D.L. & Lucchese, P.
Partner: UNT Libraries Government Documents Department

Measurement committee of the US cross section evaluation working group. Annual report, 1995

Description: The Cross Section Evaluation Working Group is a long-standing committee charged with the responsibility for organizing and overseeing the U.S. cross-section evaluation effort. It`s main product is the official U.S. evaluated nuclear data file, ENDF; the current version of this file is Version VI. All evaluations included in ENDF are reviewed and approved by CSEWG and issued by the U.S. Nuclear Data Center, Brookhaven National Laboratory. CSEWG is comprised of volunteers from the U.S. nuclear data community who possess expertise in evaluation methodologies and who collectively have been responsible for producing most of the evaluations included in ENDF. In 1992 CSEWG added the Measurements Committee to its list of standing committees and subcommittees. This was based on recognition of the importance of experimental data in the evaluation process as well as the realization that measurement activities in the U.S. were declining at an alarming rate. The mission of the Committee is to establish a network of experimentalists in the U.S. which would provide encouragement to the national nuclear data measurement effort through improved communication and facilitation of collaborative activities. The Committee currently has 19 members, and interested scientists are welcome to join the network simply by contacting the Chairman. For reference, the names of the current members and contact information are contained in this report. This annual report is the first such document issued by the Committee. It contains voluntary contributions from 10 laboratories in the U.S. which have been prepared by members of the Committee and submitted to the Chairman for compilation and editing. This report is being distributed in hard copy and is also available on-line via the National Nuclear Data Center, Brookhaven National Laboratory. It is hoped that the information provided here on the work that is going on at the reporting laboratories will prove interesting and ...
Date: August 1, 1995
Creator: Smith, D.L. & McLane, V.
Partner: UNT Libraries Government Documents Department

Impurity control in near-term tokamak reactors

Description: Several methods for reducing impurity contamination in near-term tokamak reactors by modifying the first-wall surface with a low-Z or low-sputter material are examined. A review of the sputtering data and an assessment of the technological feasibility of various wall modification schemes are presented. The power performance of a near-term tokamak reactor is simulated for various first-wall surface materials, with and without a divertor, in order to evaluate the likely effect of plasma contamination associated with these surface materials.
Date: October 1, 1976
Creator: Stacey, W. M. Jr.; Smith, D. L. & Brooks, J. N.
Partner: UNT Libraries Government Documents Department

Neutronics of a D-Li neutron source: An overview

Description: The importance of having a high energy (14 MeV) neutron source for fusion materials testing is widely recognized. The availability of a test volume with easy accessibility, with a radiation environment similar to the one expected for a fusion reactor, and with dimensions large enough to accommodate several small samples or a small blanket mock-up are requirements impossible to meet with the existing reactors and irradiation facilities. A D-Li neutron source meets the above mentioned requirements and can be built today with well known technology. This paper describes some relevant topics related to beam target configuration, neutron flux spectrum, and nuclear responses for a D-Li neutron source. The target-beam configuration is analyzed for different beam cross sectional areas and trade-offs between the area of the beam and related quantities such as available volume for testing, peak fluxes, and flux or nuclear responses gradient are presented. The conclusion is that the D-Li neutron source has the necessary characteristics to be the option of choice for IFMIF.
Date: November 1, 1993
Creator: Gomes, I. C. & Smith, D. L.
Partner: UNT Libraries Government Documents Department