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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

Fast neutron spectrum and dosimetry studies in the Coupled Fast Reactivity Measurements Facility

Description: The fast neutron spectrum of the Coupled Fast Reactivity Measurements Facility (CFRMF) at the Idaho National Engineering Laboratory (INEL) is being used to study and standardize fast reactor neutron dosimetry materials and methods. The CFRMF has been designated a ''benchmark experiment'' to test the cross section data of dosimetry materials as well as other materials used and produced in fast reactors. Information about the neutron energy spectrum of the CFRMF is presented. (auth)
Date: January 1, 1975
Creator: Rogers, J.W.; Harker, Y.D. & Millsap, D.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

CFRMF neutron radiography facility

Description: In reviewing the existing sources for neutron radiography it was determined early that the Coupled Fast Reactivity Measurement Facility (CFRMF) was the logical and economical choice. It is a swimming pool type reactor with space available for a collimator-radiograph assembly and a canal depth sufficient to reduce shielding requirements. The neutron source strength is sufficient to provide superior resolution with reasonable exposure times. The ratio of the source-sample distance to the source diameter (L/D) is an aspect ratio commonly quoted to indicate potential resolution; i.e., the greater the L/D ratio the better the resolution obtainable. Based on recent measurements using CFRMF nuclear radiograph facility, an L/D ratio of 200 has been obtained with reasonable exposure times (approximately 1 hr). The CFRMF radiograph facility is a dry tube facility. It is designed for radiographing radioactive materials which in most cases are irradiated fuel pins. It incorporates a platform placed on the canal parapet with shielding extending below the canal water level, shielded doors and positioning dowels for placement of a shielded cask. The dry tube of the radiograph assembly extends to the bottom of the canal, and into a hole extending 245 cm below the canal floor. A tapered collimater passes neutrons from the core to the samples to be radiographed. Dimensions of this collimater are: length 254 cm, maximum source aperture diameter 2.54 cm, object aperture 10 cm by 91 cm. Other source apertures of 1.9 cm and 1.27 cm are also provided. Samples longer than 91 cm require multiple radiographs to cover the entire length. Maximum sample length that can be accomodated is 416 m. Sample containment is incorporated in the positioning fixture so that contamination of the facility by the sample or vice versa is eliminated.
Date: January 1, 1978
Creator: Stepan, I.E.; Anderson, D.M. & Harker, Y.D.
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

Uncertainty Analysis of Nondestructive Assay Measurements of Nuclear Waste

Description: Regulatory agencies governing the disposal of nuclear waste require that the waste be appropriately characterized prior to disposition. The most important aspect of the characterization process, establishing radionuclide content, is often achieved by nondestructive assay (NDA). For NDA systems to be approved for use in these applications, measurement uncertainty must be established. Standard �propagation of errors� methods provide a good starting point for considering the uncertainty analysis of NDA systems for nuclear waste. However, as compared with other applications (e.g., nuclear material accountability), using NDA systems for nuclear waste measurements presents some unique challenges. These challenges, stemming primarily from the diverse nature of the waste materials encountered, carry over into the uncertainty analysis as well. This paper reviews performance measures appropriate for the assessment of NDA uncertainty, describes characteristics of nuclear waste measurements that contribute to difficulties in assessing uncertainty, and outlines some statistics based methods for incorporating variability in waste characteristics in an uncertainty analysis.
Date: November 1, 1998
Creator: Blackwood, L. G. & Harker, Y. D.
Partner: UNT Libraries Government Documents Department

Experimental investigation of filtered epithermal-photoneutron Beams for BNCT

Description: The Idaho National Engineering Laboratory (INEL) has been investigating the feasibility of a concept for an accelerator-based source of epithermal neutrons for BNCT that is based on the use of a two-stage photoneutron production process driven by an electron accelerator. In this concept, relativistic electron beams impinge upon heavily-shielded tungsten targets located at the outer radius of a small cylindrical tank of circulating heavy water (D{sub 2}0). A fraction of the energy of the electrons is converted in the tungsten targets into radially-inward-directed bremsstrahlung radiation. Neutrons subsequently generated by photodisintegration of deuterons in the D{sub 2}O within the tank are directed to the patient through a suitable beam tailoring system. Initial proof-of-principal tests using a low-current benchtop prototype of this concept have been conducted. Testing has included extensive measurements of the unfiltered photoneutron source as well as initial measurements of filtered epithermal-neutron spectra produced using two different advanced neutron filtering assemblies, as described here.
Date: December 31, 1996
Creator: Nigg, D.W.; Mitchell, H.E.; Harker, Y.D. & Harmon, J.F.
Partner: UNT Libraries Government Documents Department

SWEPP PAN assay system uncertainty analysis: Active mode measurements of solidified aqueous sludge waste

Description: The Idaho National Engineering and Environmental Laboratory is being used as a temporary storage facility for transuranic waste generated by the US Nuclear Weapons program at the Rocky Flats Plant (RFP) in Golden, Colorado. Currently, there is a large effort in progress to prepare to ship this waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Active Neutron (PAN) radioassay system. This paper is one of a series of reports quantifying the results of the uncertainty analysis of the PAN system measurements for specific waste types and measurement modes. In particular this report covers active mode measurements of weapons grade plutonium-contaminated aqueous sludge waste contained in 208 liter drums (item description codes 1, 2, 7, 800, 803, and 807). Results of the uncertainty analysis for PAN active mode measurements of aqueous sludge indicate that a bias correction multiplier of 1.55 should be applied to the PAN aqueous sludge measurements. With the bias correction, the uncertainty bounds on the expected bias are 0 {+-} 27%. These bounds meet the Quality Assurance Program Plan requirements for radioassay systems.
Date: December 1, 1997
Creator: Blackwood, L.G.; Harker, Y.D. & Meachum, T.R.
Partner: UNT Libraries Government Documents Department

Material Identification Technology (MIT) concept technical feasibility study

Description: The Idaho National Engineering Laboratory (INEL) has initiated the design and development of a novel pulsed accelerator-based, active interrogation concept. The proposed concept, referred to as the Material Identification Technology (MIT), enables rapid (between accelerator pulses), non-destructive, elemental composition analysis of both nuclear and non-nuclear materials. Applications of this technique include material monitoring in support of counter-proliferation activities, such as export controls (at domestic and international inspection locations), SNM controls, nuclear weapon dismantlement, and chemical weapon verification. Material Identification Technology combines a pulsed, X-ray source (an electron accelerator) and a gamma detection system. The accelerator must maximize neutron production (pulse width, beam current, beam energy, and repetition rate) and minimize photon dose to the object. Current available accelerator technology can meet these requirements. The detection system must include detectors which provide adequate gamma energy resolution capability, rapid recovery after the initial X-ray interrogation pulse, and multiple single gamma event detection between accelerator pulses. Further research is required to develop the detection system. This report provides the initial feasibility assessment of the MIT concept.
Date: September 1, 1993
Creator: Jones, J.L.; Harker, Y.D.; Yoon, W.Y. & Johnson, L.O.
Partner: UNT Libraries Government Documents Department

Pulsed photoneutron interrogation: The GNT demonstration system

Description: The Idaho National Engineering Laboratory (INEL) has developed and tested an active photon interrogation technique to support the Department of Energy`s (DOE) Office of National Security and Nonproliferation (NN) mission related to verification technologies development. The INEL concept, referred to as the Gamma-Neutron Threshold (GNT) technology, uses a transportable, field-deployable, selective-energy (2 to 10 MeV), pulsed, electron accelerator to produce energetic X-rays having a bremsstrahlung spectrum. The energetic X-rays induce neutrons in many proliferation-limited items via direct photoneutron/photofission interactions. The time-dependent neutron response, as a function of the electron beam energy, is measured with a tripod-mounted, detector assembly and a portable data acquisition system. The portable detector assembly has been specifically designed to operate in very intense, pulsed X-ray environments. The GNT technique measures both the prompt and delayed neutron emission after each accelerator pulse. This report fully describes each component of this system and presents various signature results based on these emissions.
Date: October 1, 1994
Creator: Jones, J.L.; Harker, Y.D.; Yoon, W.Y.; Hoggan, J.M. & McManus, G.J.
Partner: UNT Libraries Government Documents Department

Destructive versus Nondestructive Assay Comparisons Using the SWEPP Gamma-Ray Spectrometer

Description: In support of data quality objectives for the INEEL Stored Waste Examination Pilot Plant (SWEPP) a series of 208-liter (55-gallon) waste drums containing inorganic sludge have been sampled and destructively analyzed. The drums were non-destructively assayed by the SWEPP PAN system and the SWEPP Gamma-Ray Spectrometer (SGRS) prior to sampling. This paper reports some of the conclusions from the destructive versus NDA comparisons, and additionally presents the results of an on-going effort to use the destructive analyses to validate absolute efficiency curves calculated using Monte-Carlo and analytical modeling for the SGRS. Destructive analysis results are available from radiochemical assay of 128 sludge-containing drums. The content codes represented are CC001 (42 items), CC002 (8), CC007 (48), CC800 (16), CC803 (3), and CC807 (11.) Each drum had two full-length vertical cores removed from designated radial positions. The entire length of each core was composited and submitted for analysis. All of the core composites were analyzed radiochemically for Am-241, Pu-239/240, and Pu-238, and by inductively-coupled mass spectrometry (ICPMS) for U-235 and U-238. Not only have the destructive analysis results been useful in documenting the performance of both the SGRS and the PAN system, but also have allowed the determination of certain absolute counting efficiency values for the SGRS. The values, in turn will allow us to validate SGRS counting efficiencies computed by MCNP and analytical modeling, and perhaps use the SGRS as an absolute assay technique.
Date: November 1, 1998
Creator: Killian, E. W.; Hartwell, J. K.; Yoon, W. & Harker, Y. D.
Partner: UNT Libraries Government Documents Department

Technical specifications manual for the MARK-1 pulsed ionizing radiation detection system. Volume 1

Description: The MARK-1 detection system was developed by the Idaho National Engineering Laboratory for the US Department of Energy Office of Arms Control and Nonproliferation. The completely portable system was designed for the detection and analysis of intense photon emissions from pulsed ionizing radiation sources. This manual presents the technical design specifications for the MARK-1 detection system and was written primarily to assist the support or service technician in the service, calibration, and repair of the system. The manual presents the general detection system theory, the MARK-1 component design specifications, the acquisition and control software, the data processing sequence, and the system calibration procedure. A second manual entitled: Volume 2: Operations Manual for the MARK-1 Pulsed Ionizing Radiation Detection System (USDOE Report WINCO-1108, September 1992) provides a general operational description of the MARK-1 detection system. The Operations Manual was written primarily to assist the field operator in system operations and analysis of the data.
Date: March 1, 1993
Creator: Lawrence, R. S.; Harker, Y. D.; Jones, J. L. & Hoggan, J. M.
Partner: UNT Libraries Government Documents Department

Uncertainty analysis of the SWEPP PAN assay system for glass waste (content codes 440, 441 and 442)

Description: INEL is being used as a temporary storage facility for transuranic waste generated by the Nuclear Weapons program at the Rocky Flats Plant. Currently, there is a large effort in progress to prepare to ship this waste to WIPP. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Action Neutron (PAN) radioassay system. This paper discusses a modified statistical sampling and verification approach used to determine the total uncertainty of SWEPP PAN measurements for glass waste (content codes 440, 441, and 442) contained in 208 liter drums. In the modified statistical sampling and verification approach, the total performance of the SWEPP PAN nondestructive assay system for specifically selected waste conditions is simulated using computer models. A set of 100 cases covering the known conditions exhibited in glass waste was compiled using a combined statistical sampling and factorial experimental design approach. Parameter values assigned in each simulation were derived from reviews of approximately 100 real-time radiography video tapes of RFP glass waste drums, results from previous SWEPP PAN measurements on glass waste drums, and shipping data from RFP where the glass waste was generated. The data in the 100 selected cases form the multi-parameter input to the simulation model. The reported plutonium masses from the simulation model are compared with corresponding input masses. From these comparisons, the bias and total uncertainty associated with SWEPP PAN measurements on glass waste drums are estimated. The validity of the simulation approach is verified by comparing simulated output against results from calibration measurements using known plutonium sources and two glass waste calibration drums.
Date: October 1, 1996
Creator: Blackwood, L.G.; Harker, Y.D.; Meachum, T.R. & Yoon, W.Y.
Partner: UNT Libraries Government Documents Department

INEL test plan for evaluating waste assay systems

Description: A test bed is being established at the Idaho National Engineering Laboratory (INEL) Radioactive Waste Management Complex (RWMC). These tests are currently focused on mobile or portable radioassay systems. Prior to disposal of TRU waste at the Waste Isolation Pilot Plant (WIPP), radioassay measurements must meet the quality assurance objectives of the TRU Waste Characterization Quality Assurance Program Plan. This test plan provides technology holders with the opportunity to assess radioassay system performance through a three-tiered test program that consists of: (a) evaluations using non-interfering matrices, (b) surrogate drums with contents that resemble the attributes of INEL-specific waste forms, and (c) real waste tests. Qualified sources containing a known mixture and range of radionuclides will be used for the non-interfering and surrogate waste tests. The results of these tests will provide technology holders with information concerning radioassay system performance and provide the INEL with data useful for making decisions concerning alternative or improved radioassay systems that could support disposal of waste at WIPP.
Date: September 1, 1996
Creator: Mandler, J.W.; Becker, G.K.; Harker, Y.D.; Menkhaus, D.E. & Clements, T.L. Jr.
Partner: UNT Libraries Government Documents Department

Passive active neutron radioassay measurement uncertainty for combustible and glass waste matrices

Description: Using a modified statistical sampling and verification approach, total uncertainty of INEL`s Passive Active Neutron (PAN) radioassay system was evaluated for combustible and glass content codes. Waste structure and content of 100 randomly selected drums in each the waste categories were computer modeled based on review of real-time radiography video tapes. Specific quantities of Pu were added to the drum models according to an experimental design. These drum models were then submitted to the Monte Carlo Neutron Photon code processing and subsequent calculations to produce simulated PAN system measurements. The reported Pu masses from the simulation runs were compared with the corresponding input masses. Analysis of the measurement errors produced uncertainty estimates. This paper presents results of the uncertainty calculations and compares them to previous reported results obtained for graphite waste.
Date: January 1, 1997
Creator: Blackwood, L.G.; Harker, Y.D.; Meachum, T.R. & Yoon, Woo Y.
Partner: UNT Libraries Government Documents Department

Validation of computational methods for treatment planning of fast-neutron therapy using activation foil techniques

Description: A closed-form direct method for unfolding neutron spectra from foil activation data is presented. The method is applied to measurements of the free-field neutron spectrum produced by the proton-cyclotron-based fast-neutron radiotherapy facility at the University of Washington (UW) School of Medicine. The results compare favorably with theoretical expectations based on an a-priori calculational model of the target and neutron beamline configuration of the UW facility.
Date: December 1, 1997
Creator: Nigg, D.W.; Wemple, C.A.; Hartwell, J.K.; Harker, Y.D.; Venhuizen, J.R. & Risler, R.
Partner: UNT Libraries Government Documents Department

SWEPP PAN assay system uncertainty analysis: Passive mode measurements of graphite waste

Description: The Idaho National Engineering and Environmental Laboratory is being used as a temporary storage facility for transuranic waste generated by the U.S. Nuclear Weapons program at the Rocky Flats Plant (RFP) in Golden, Colorado. Currently, there is a large effort in progress to prepare to ship this waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Active Neutron (PAN) radioassay system. To this end a modified statistical sampling and verification approach has been developed to determine the total uncertainty of a PAN measurement. In this approach the total performance of the PAN nondestructive assay system is simulated using computer models of the assay system and the resultant output is compared with the known input to assess the total uncertainty. This paper is one of a series of reports quantifying the results of the uncertainty analysis of the PAN system measurements for specific waste types and measurement modes. In particular this report covers passive mode measurements of weapons grade plutonium-contaminated graphite molds contained in 208 liter drums (waste code 300). The validity of the simulation approach is verified by comparing simulated output against results from measurements using known plutonium sources and a surrogate graphite waste form drum. For actual graphite waste form conditions, a set of 50 cases covering a statistical sampling of the conditions exhibited in graphite wastes was compiled using a Latin hypercube statistical sampling approach.
Date: July 1, 1997
Creator: Blackwood, L. G.; Harker, Y. D.; Meachum, T. R. & Yoon, Woo Y.
Partner: UNT Libraries Government Documents Department

Proof-of-Concept Assessment of a Photofission-Based Interrogation System for the Detection of Shielded Nuclear Material

Description: A photonuclear interrogation method was experimentally assessed for the detection of shielded nuclear materials. Proof-of-Concept assessment was performed at the Los Alamos National Laboratory (LANL) TA-18 facility and used the INEEL VARITRON electron accelerator. Experiments were performed to assess and characterize the delayed neutron emission responses for different nuclear materials with various shield configurations using three ''nominal'' electron beam energies; 8-, 10-, and 11-MeV. With the exception of highly enriched uranium (HEU), the nuclear materials assessed represent material types commonly encountered in commerce. The specific nuclear materials studied include a solid 4.8-kg HEU sphere, a 5-kg multiple-object, depleted uranium (DU) [uranium with about 0.2% enrichment with U-235] target, and two 11-kg thorium disks. The shield materials selected include polyethylene, borated-polyethylene, and lead. Experimental results, supported with numerical predictions, have shown that the photonuclear interrogation technique is quite capable of detecting shielded nuclear material via the direct measurement of the photofission-induced delayed neutron emissions. To identify or discriminate between nuclear material types (i.e., depleted uranium, HEU, and thorium), a ratio of delayed neutron counts at two different beam energies is utilized. This latter method, referred to as the dual-beam energy ratio Figure-of-Merit, allows one to differentiate among the three nuclear material types.
Date: November 1, 2000
Creator: Jones, J. L.; Yoon, W. Y.; Harker, Y. D.; Hoggan, J. M.; Haskell, K. J. & VanAusdeln, L. A.
Partner: UNT Libraries Government Documents Department