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Actinide integral measurements in the CFRMF and integral tests for ENDF/B-V

Description: Integral capture and/or fission rates have been reported earlier for several actinides irradiated in the fast neutron field of the Coupled Fast Reactivity Measurements Facility (CFRMF). These nuclides include /sup 232/Th, /sup 233/U, /sup 235/U, /sup 238/U, /sup 237/Np, /sup 239/Pu, /sup 240/Pu, /sup 242/Pu, /sup 241/Am and /sup 243/Am. This paper forucses on the utilization of these integral data for testing the respective cross sections on ENDF/B-V. Integral cross sections derived from the measured reaction rates are tabulated. Results are presented for cross-section data testing which includes integral testing based on a comparison of calculated and measured integral cross sections and testing based on least-squares-adjustment analyses.
Date: January 1, 1982
Creator: Anderl, R.A.
Partner: UNT Libraries Government Documents Department

Cross-section evaluation utilizing integral reaction-rate measurements in fast neutron fields

Description: The role of integral reaction-rate data for cross-section evaluation is reviewed. The subset of integral data considered comprises integral reaction rates measured for dosimeter, fission-product, and actinide-type materials irradiated in reactor dosimetry fast neutron benchmark fields and in the EBR-II. Utilization of these integral data for integral testing, multigroup cross-section adjustment and pointwise cross section adjustment is treated in some detail. Examples are given that illustrate the importance of considering a priori uncertainty and correlation information for these analyses. 3 figures, 3 tables.
Date: January 1, 1980
Creator: Anderl, R.A.
Partner: UNT Libraries Government Documents Department

Laboratory-scale shielded cell for /sup 252/Cf

Description: A shielded-cell facility for storing and handling remotely up to 2 milligram quantities of unencapsulated /sup 252/Cf has been built in a radiochemistry laboratory at the Test Reactor Area of the Idaho National Engineering Laboratory. Unique features of this facility are its compact bulk radiation shield of borated gypsum and transfer lines which permit the transport of fission product activity from /sup 252/Cf fission sources within the cell to a mass separator and to a fast radiochemistry system in nearby rooms.
Date: January 1, 1979
Creator: Anderl, R.A. & Cargo, C.H.
Partner: UNT Libraries Government Documents Department

Measurement of the integral capture and fission cross sections for /sup 232/Th in the CFRMF

Description: The recent evaluation of the cross-section data bases for /sup 232/Th capture and fission emphasized significant normalization discrepancies between the available differential data. To help resolve the normalization discrepancies, the capture and fission integral cross sections were measured for /sup 232/Th in the fast neutron zone of the Coupled Fast Reactivity Measurements Facility (CFRMF). The cross sections are derived from the radiometric determination of the saturation reaction rates for fission and capture based on the Ge(Li) spectrometric measurement of the absolute gamma emission rates of the 537-keV and 1596-keV lines in the /sup 140/Ba - /sup 140/La decay and the 311.9-keV line in the /sup 233/Pa decay. For capture and fission, respectively, the measured integral cross sections are 291 mb +- 3% and 19.6 mb +- 5%. The ratios of the integral cross sections computed with ENDF/B-IV thorium cross sections and the CFRMF neutron spectrum to the above values are 0.99 for capture and 0.90 for fission. 19 references.
Date: January 1, 1979
Creator: Anderl, R.A. & Harker, Y.D.
Partner: UNT Libraries Government Documents Department

ATR neutron spectral characterization

Description: The Advanced Test Reactor (ATR) at INEL provides intense neutron fields for irradiation-effects testing of reactor material samples, for production of radionuclides used in industrial and medical applications, and for scientific research. Characterization of the neutron environments in the irradiation locations of the ATR has been done by means of neutronics calculations and by means of neutron dosimetry based on the use of neutron activation monitors that are placed in the various irradiation locations. The primary purpose of this report is to present the results of an extensive characterization of several ATR irradiation locations based on neutron dosimetry measurements and on least-squares-adjustment analyses that utilize both neutron dosimetry measurements and neutronics calculations. This report builds upon the previous publications, especially the reference 4 paper. Section 2 provides a brief description of the ATR and it tabulates neutron spectral information for typical irradiation locations, as derived from the more historical neutron dosimetry measurements. Relevant details that pertain to the multigroup neutron spectral characterization are covered in section 3. This discussion includes a presentation on the dosimeter irradiation and analyses and a development of the least-squares adjustment methodology, along with a summary of the results of these analyses. Spectrum-averaged cross sections for neutron monitoring and for displacement-damage prediction in Fe, Cr, and Ni are given in section 4. In addition, section4 includes estimates of damage generation rates for these materials in selected ATR irradiation locations. In section 5, the authors present a brief discussion of the most significant conclusions of this work and comment on its relevance to the present ATR core configuration. Finally, detailed numerical and graphical results for the spectrum-characterization analyses in each irradiation location are provided in the Appendix.
Date: November 1, 1995
Creator: Rogers, J.W. & Anderl, R.A.
Partner: UNT Libraries Government Documents Department

Neodymium, samarium and europium capture cross-section adjustments based on EBR-II integral measurements

Description: Integral capture measurements were made for highly enriched isotopes of neodymium, samarium, and europium irradiated in a row 8 position of EBR-II with samples located both at mid-plane and in the axial reflector. Broad response, resonance, and threshold dosimeters were included to characterize the neutron spectra at the sample locations. The saturation reaction rates for the rare-earth samples were determined by post-irradiation mass-spectrometric analyses and for the dosimeter materials by the gamma-spectrometric method. The HEDL maximum-likelihood analysis code, FERRET, was used to make a least-squares adjustment of the ENDF/B-IV rare-earth cross sections based on the measured dosimeter and fission-product reaction rates. Preliminary results to date indicate a need for a significant upward adjustment of the capture cross sections for /sup 143/Nd, /sup 145/Nd, /sup 147/Sm, and /sup 148/Sm. 8 figures, 2 tables.
Date: January 1, 1979
Creator: Anderl, R.A.; Harker, Y.D. & Schmittroth, F.
Partner: UNT Libraries Government Documents Department

Basic studies of a gas-jet-coupled ion source for on-line isotope separation

Description: A hollow-cathode ion source was used in a gas-jet-coupled configuration to produce ion beams of fission products transported to it from a /sup 252/Cf fission source. Solid aerosols of NaCl and Ag were used effectively as activity carriers in the gas-jet system. Flat-plate skimmers provided an effective coupling of the ion source to the gas jet. Ge(Li) spectrometric measurements of the activity deposited on an ion-beam collector relative to that deposited on a pre-skimmer collector were used to obtain separation efficiencies ranging from 0.1% to > 1% for Sr, Y, Tc, Te, Cs, Ba, Ce, Pr, Nd and Sm. The use of CCl/sub 4/ as a support gas resulted in a significant enhancement of the alkaline-earth and rare-earth separation efficiencies.
Date: January 1, 1980
Creator: Anderl, R.A.; Novick, V.J. & Greenwood, R.C.
Partner: UNT Libraries Government Documents Department

Integral capture cross-section measurements in the CFRMF for LMFBR control materials

Description: Integral capture-cross sections for separated isotopes of Eu and Ta are reported for measurements in the Coupled Fast Reactivity Measurements Facility (CFRMF). These cross sections along with that measured in the CFRMF for $sup 10$B(n,$alpha$) provide an absolute standard for evaluating the relative reactivity worth of Eu$sub 2$O$sub 3$, B$sub 4$C and Ta in neutron fields typical of an LMFBR core. Based on these measurements and for neutron fields characterized by the $sup 235$U:$sup 238$U reaction rate spectral index ranging from 23 to 50, the infinitely dilute relative worth of Eu$sub 2$O$sub 3$ has been estimated to be 25 to 40 percent higher than that for B$sub 4$C and 80 percent to 100 percent higher than that for Ta. 11 references. (auth)
Date: January 1, 1975
Creator: Anderl, R.A.; Harker, Y.D.; Turk, E.H.; Nisle, R.G. & Berreth, J.R.
Partner: UNT Libraries Government Documents Department

Integral measurements for higher actinides in CFRMF. [0. 1 to 2000 keV]

Description: To improve upon the lack of fast integral data for higher actinides, an effort is underway to measure integral capture and fission cross sections for /sup 242/Pu, /sup 241/Am and /sup 243/Am in the fast neutron zone of the Couple Fast Reactivity Measurements Facility (CFRMF). Fission cross sections are determined based on the Ge(Li) gamma spectrometric measurements of the absolute emission rates of the 537-keV and/or 1596-keV lines in the /sup 140/Ba - /sup 140/La decay. The capture rate for /sup 242/Pu is based on the measurement of the absolute emission rate of the 84.0 keV line in the /sup 243/Pu ..beta../sup -/ decay. Although the capture cross sections for /sup 241/Am and /sup 243/Am are not obtained directly, the cross sections for production of /sup 242/Cm and /sup 244/Cm are based on the quantitative alpha spectrometry and total alpha counting. Measured integral and capture cross sections for /sup 242/Pu are 357 mb +- 10% and 146 mb +- 15%. Corresponding spectral averaged cross sections calculated using ENDF/B-IV data and 489 mb and 238 mb, respectively. For /sup 241/Am fission and capture the measured cross sections are 504 mb +- 12% and 1.01 b +- 3%, respectively. For /sup 243/Am fission and capture, the measure cross sections are 0.352 b and .10 b, respectively. 19 references.
Date: January 1, 1979
Creator: Harker, Y.D.; Anderl, R.A.; Turk, E.H. & Schroeder, N.C.
Partner: UNT Libraries Government Documents Department

Fusion Safety Program Annual Report, Fiscal Year 1996

Description: This report summarizes the major activities of the Fusion Safety Program in FY 1996. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory, and Lockheed Martin Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. The objective is to perform research and develop data needed to ensure safety in fusion facilities. Activities include experiments, analysis, code development and application, and other forms of research. These activities are conducted at the INEL, at other DOE laboratories, and at other institutions. Among the technical areas covered in this report are tritium safety, chemical reactions and activation product release, risk assessment failure rate database development, and safety code development and application to fusion safety issues. Most of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Work done for ITER this year has focused on developing the needed information for the Non- Site- Specific Safety Report (NSSR-1). A final area of activity described is development of the new DOE Technical Standards for Safety of Magnetic Fusion Facilities.
Date: December 1, 1996
Creator: Longhurst, G.R.; Anderl, R.A. & Cadwallader, L.C.
Partner: UNT Libraries Government Documents Department

Tritium loading in ITER plasma-facing surfaces and its release under accident conditions

Description: Plasma-facing surfaces of the International Thermonuclear Experimental Reactor (ITER) will take up tritium from the plasma. These surfaces will probably consist of matures of Be, C, and possibly W together with other impurities. Recent experimental results have suggested mechanisms, not previously considered in analyses, by which tritium and other hydrogen isotopes are retained in Be. This warrants revised modeling and estimation of the amount of tritium that will be deposited in ITER beryllium plasma-facing surfaces and the rates at which it can be released under postulated accident scenarios. In this paper we describe improvements in modeling and experiments planned at the Idaho National Engineering Laboratory (INEL) to investigate the tritium uptake and thermal release behavior for mixed plasma- facing materials. TMAP4 calculations were made using recent data to estimate first-wall tritium inventories in ITER. 16 refs., 1 fig.
Date: September 1, 1996
Creator: Longhurst, G.R.; Anderl, R.A. & Pawelko, R.J.
Partner: UNT Libraries Government Documents Department

Development of an IEC neutron source for NDE

Description: This paper concerns the development of a neutron so based on the inertial electrostatic confinement (IEC) of a low density fusion plasma in a gridded, spherically-focusing device. With the motivation of using such sources for nondestructive evaluation (NDE) applications, the focus of the development is on : Small size devices, sealed operation with D{sub 2} or D{sub 2}/T{sub 2} mixtures, Power-utilization and neutron-output optimization, and integration into an assay system. In this paper, we describe an experimental system that has been established for the development and testing of IEC neutron sources, and we present preliminary results of tests conducted for 25-cm and 15-cm diameter IEC devices.
Date: December 1, 1995
Creator: Anderl, R.A,; Hartwell, J.K. & Nadler, J.H.
Partner: UNT Libraries Government Documents Department

Neutron spectrum studies in the ATR (Advanced Test Reactor)

Description: The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) has been and currently is used to provide irradiation fields to study the effects of intense radiation on samples of reactor materials. These samples include fuel, cladding, control and structural materials. The ATR is also used to irradiate target materials for the production of radionuclides used in industrial and medical applications as well as for scientific research. Routine monitoring of the thermal'' and fast'' neutron levels have been conducted during every operational cycle since its startup in 1970. The routine neutron dosimetry has been primarily accomplished using the {sup 59}Co(n,{gamma}){sup 60}Co reaction for thermal'' neutrons and the {sup 58}Ni(n,p) {sup 58}Co reaction for fast'' neutrons as described in ASTM standard methods E261, E262, and E264. Neutron spectrum studies have now been conducted in the epithermal and fast neutron energy ranges for the various capsule irradiation test facilities and the routine neutron monitoring locations. 7 refs., 5 figs., 1 tab.
Date: January 1, 1990
Creator: Rogers, J.W.; Anderl, R.A. & Putnam, M.H.
Partner: UNT Libraries Government Documents Department

Fusion safety program annual report fiscal year 1997

Description: This report summarizes the major activities of the Fusion Safety Program in FY 1997. The Idaho National Engineering and Environmental Laboratory (INEEL) is the designated lead laboratory, and Lockheed Martin Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in FY 1979 to perform research and develop data needed to ensure safety in fusion facilities. Activities include experiments, analysis, code development and application, and other forms of research. These activities are conducted at the INEEL, different DOE laboratories, and other institutions. The technical areas covered in this report include chemical reactions and activation product release, tritium safety, risk assessment failure rate database development, and safety code development and application to fusion safety issues. Most of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER) project. Work done for ITER this year has focused on developing the needed information for the Non-site Specific Safety Report (NSSR-2).
Date: January 1, 1998
Creator: Longhurst, G.R.; Anderl, R.A. & Cadwallader, L.C.
Partner: UNT Libraries Government Documents Department

In-vessel tritium retention and removal in ITER

Description: The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews ...
Date: June 1998
Creator: Federici, G.; Anderl, R. A. & Andrew, P.
Partner: UNT Libraries Government Documents Department

Steam chemical reactivity of plasma-sprayed beryllium

Description: Plasma-spraying with the potential for in-situ repair makes beryllium a primary candidate for plasma facing and structural components in experimental magnetic fusion machines. Deposits with good thermal conductivity and resistance to thermal cycling have been produced with low pressure plasma-spraying (LPPS). A concern during a potential accident with steam ingress is the amount of hydrogen produced by the reactions of steam with hot components. In this study the authors measure the reaction rates of various deposits produced by LPPS with steam from 350 C to above 1,000 C. They correlate these reaction rates with measurements of density, open porosity and BET surface areas. They find the reactivity to be largely dependent upon effective surface area. Promising results were obtained below 600 C from a 94% theoretical dense (TD) deposit with a BET specific surface area of 0.085 m{sup 2}/g. Although reaction rates were higher than those for dense consolidated beryllium they were substantially lower, i.e., about two orders of magnitude, than those obtained from previously tested lower density plasma-sprayed deposits.
Date: July 1, 1998
Creator: Anderl, R.A.; Pawelko, R.J.; Smolik, G.R. & Castro, R.G.
Partner: UNT Libraries Government Documents Department

Sensitivity and a priori uncertainty analysis of the CFRMF central flux spectrum

Description: The Coupled Fast Reactivity Measurements Facility (CFRMF), located at the Idaho National Engineering Laboratory, is a zoned-core critical assembly with a fast-neutron-spectrum zone in the center of an enriched /sup 235/U, water-moderated thermal driver. An accurate knowledge of the central neutron spectrum is important to data-testing analyses which utilize integral reaction-rate data measured for samples placed in the CFRMF field. The purpose of this paper is to present the results of a study made with the AMPX-II and FORSS code systems to deterine the central-spectrum flux covariance matrix due to uncertainties and correlations in the nuclear data for the materials which comprise the facility.
Date: January 1, 1980
Creator: Ryskamp, J.M.; Anderl, R.A.; Broadhead, B.L.; Ford, W.E. III; Lucius, J.L. & Marable, J.H.
Partner: UNT Libraries Government Documents Department

The Safety and Tritium Applied Research (STAR) Facility: Status-2004*

Description: The purpose of this paper is to present the current status of the development of the Safety and Tritium Applied Research (STAR) Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). Designated a National User Facility by the US DOE, the primary mission of STAR is to provide laboratory infrastructure to study tritium science and technology issues associated with the development of safe and environmentally friendly fusion energy. Both tritium and non-tritium fusion safety research is pursued along three key thrust areas: (1) plasma-material interactions of plasma-facing component (PFC) materials exposed to energetic tritium and deuterium ions, (2) fusion safety concerns related to PFC material chemical reactivity and dust/debris generation, activation product mobilization, and tritium behavior in fusion systems, and (3) molten salts and fusion liquids for tritium breeder and coolant applications. STAR comprises a multi-room complex with operations segregated to permit both tritium and non-tritium activities in separately ventilated rooms. Tritium inventory in STAR is limited to 15,000 Ci to maintain its classification as a Radiological Facility. Experiments with tritium are typically conducted in glovebox environments. Key components of the tritium infrastructure have been installed and tested. This includes the following subsystems: (1) a tritium Storage and Assay System (SAS) that uses two 50-g depleted uranium beds for tritium storage and PVT/beta-scintillation analyses for tritium accountability measurements, (2) a Tritium Cleanup System (TCS) that uses catalytic oxidation and molecular sieve water absorption to remove tritiated species from glovebox atmosphere gases and gaseous effluents from experiment and process systems, and (3) tritium monitoring instrumentation for room air, glovebox atmosphere and stack effluent tritium concentration measurements. Integration of the tritium infrastructure subsystems with the experimental and laboratory process systems is planned for early in 2004. Following an operational readiness review, tritium operations will be initiated in the summer of 2004. ...
Date: September 1, 2004
Creator: Anderl, R. A.; Longhurst, G. R.; Pawelko, R. J.; Sharpe, J. P.; Schuetz, S. T. & Petti, D. A.
Partner: UNT Libraries Government Documents Department

Tokamak dust particle size and surface area measurement

Description: The INEEL has analyzed a variety of dust samples from experimental tokamaks: General Atomics` DII-D, Massachusetts Institute of Technology`s Alcator CMOD, and Princeton`s TFTR. These dust samples were collected and analyzed because of the importance of dust to safety. The dust may contain tritium, be activated, be chemically toxic, and chemically reactive. The INEEL has carried out numerous characterization procedures on the samples yielding information useful both to tokamak designers and to safety researchers. Two different methods were used for particle characterization: optical microscopy (count based) and laser based volumetric diffraction (mass based). Surface area of the dust samples was measured using Brunauer, Emmett, and Teller, BET, a gas adsorption technique. The purpose of this paper is to present the correlation between the particle size measurements and the surface area measurements for tokamak dust.
Date: July 1, 1998
Creator: Carmack, W.J.; Smolik, G.R.; Anderl, R.A.; Pawelko, R.J. & Hembree, P.B.
Partner: UNT Libraries Government Documents Department

Beryllium Interactions in Molten Salts

Description: Molten flibe (2LiF·BeF2) is a candidate as a cooling and tritium breeding media for future fusion power plants. Neutron interactions with the salt will produce tritium and release excess free fluorine ions. Beryllium metal has been demonstrated as an effective redox control agent to prevent free fluorine, or HF species, from reacting with structural metal components. The extent and rate of beryllium solubility in a pot design experiments to suppress continuously supplied hydrogen fluoride gas has been measured and modeled[ ]. This paper presents evidence of beryllium loss from specimens, a dependence of the loss upon bi-metal coupling, i.e., galvanic effect, and the partitioning of the beryllium to the salt and container materials. Various posttest investigative methods, viz., scanning electron microscopy (SEM), Auger electron spectroscopy (AES) and x-ray photoelectron spectroscopy (XPS) were used to explore this behavior.
Date: January 1, 2006
Creator: Smolik, G. S.; Simpson, M. F.; Pinhero, P. J.; Hara, M.; Hatano, Y.; Anderl, R. A. et al.
Partner: UNT Libraries Government Documents Department

JUPITER-II Molten Salt Flibe Research: An Update On Tritium, Mobilization and Redox Chemistry Experiments

Description: The second Japan/US Program on Irradiation Tests for Fusion Research (JUPITER-II) began on April 1, 2001. Part of the collaborative research centers on studies of the molten salt 2LiF2–BeF2 (also known as Flibe) for fusion applications. Flibe has been proposed as a self-cooled breeder in both magnetic and inertial fusion power plant designs over the last 25 years. The key feasibility issues associated with the use of Flibe are the corrosion of structural material by the molten salt, tritium behavior and control in the molten salt blanket system, and safe handling practices and releases from Flibe during an accidental spill. These issues are all being addressed under the JUPITER-II program at the Idaho National Laboratory in the Safety and Tritium Applied Research (STAR) facility. In this paper, we review the program to date in the area of tritium/deuterium behavior, Flibe mobilization under accident conditions and testing of Be as a redox agent to control corrosion. Future activities planned through the end of the collaboration are also presented.
Date: May 1, 2005
Creator: Petti, D.A.; Petti, D. A.; Smolik, G. R.; Simpson, Michael F.; Sharpe, John P.; Anderl, R. A. et al.
Partner: UNT Libraries Government Documents Department