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Self-Shielding Measurements in PPA-20

Description: From Introduction: "The self-shielding characteristics of a number of absorbers and of U-235 and U-238 have been investigated in PPA-20 (66°). Most of the measurements have been made in Ring 3 at the core mid-plane. The foil doubling technique was employed. Particular attention was given to 'thin' foil regions for the purpose of extrapolating the data to zero thickness."
Date: September 9, 1954
Creator: King, J. S.
Partner: UNT Libraries Government Documents Department

Integrated Box Interrogation System (IBIS) Preliminary Design Study

Description: Canberra Industries has won the tendered solicitation, INEEL/EST-99-00121 for boxed waste Nondestructive Assay Development and Demonstration. Canberra will provide the Integrated Box Interrogation System (IBIS) which is a suite of assay instrumentation and a data reduction system that addresses the measurement needs for Boxed Wastes identified in the solicitation and facilitates the associated experimental program and demonstration of system capability. The IBIS system will consist of the next generation CWAM system, i.e. CWAM II, which is a Scanning Passive/Active Neutron interrogation system which we will call a Box Segmented Neutron Scanner (BSNS), combined with a physically separate Box Segmented Gamma-ray Scanning (BSGS) system. These systems are based on existing hardware designs but will be tailored to the large sample size and enhanced to allow the program to evaluate the following measurement criteria:Characterization and correction for matrix heterogeneity Characterization of non-uniform radio-nuclide and isotopic compositions Assay of high density matrices (both high-Z and high moderator contents)Correction for radioactive material physical form - such as self shielding or multiplication effects due to large accumulations of radioactive materials.Calibration with a minimal set of reference standards and representative matrices.THis document summarizes the conceptual design parameters of the IBIS and indicates areas key to the success of the project where development is to be centered. The work presented here is a collaborative effort between scientific staff within Canberra and within the NIS-6 group at LANL.
Date: January 13, 2003
Creator: Croft, Stephen; Martancik, David; Young, Brian; Chard, Patrick M. J.; Estop, Robert J.; Melton, Sheila et al.
Partner: UNT Libraries Government Documents Department


Description: The requirements and performance goals for the collimators are to reduce the uncontrolled beam loss by 2 x 10{sup -4} absorb 2 kW of deposited heat, and minimize production and leakage of secondary radiation. In order to meet these requirements a self-shielding collimator configuration consisting of a layered structure was designed. The front layers (in the direction of the proton beam) are relatively transparent to the protons, and become progressively less transparent (blacker) with depth into the collimator. In addition, a high density (iron) shield is added around the outside. The protons will be stopped in the center of the collimator, and thus the bulk of the secondary particles are generated at this location. The conceptual design described, the method of analysis discussed, and preliminary performance parameters outlined.
Date: March 29, 1999
Partner: UNT Libraries Government Documents Department

Verification of High Temperature Free Atom Thermal Scattering in MERCURY Compared to TART

Description: This is part of a series of reports verifying the accuracy of the relatively new MERCURY [1] Monte Carlo particle transport code by comparing its results to those of the older TART [2] Monte Carlo particle transport code. In the future we hope to extend these comparisons to include deterministic (Sn) codes [3]. Here we verify the accuracy of the free atom thermal scattering model [4] by using it over a very large temperature range. We would like to be able to use these Monte Carlo codes for astrophysical applications, where the temperature of the medium can be extremely high compared to the temperatures we normally encounter in our terrestrial applications [5]. The temperature is so high that is it often defined in eV rather than Kelvin. For a correspondence between the two scale 293.6 Kelvin (room temperature) corresponds to 0.0253 eV {approx} 1/40 eV. So that 1 eV temperature is about 12,000 Kelvin, and 1 keV temperature is about 12 million Kelvin. Here we use a relatively small system measured in cm, but by using {rho}R scaling [6] our results are equally applicable to systems measured in Km or thousands of Km or any size that we need for astrophysical applications. The emphasis here is not on modeling any given real system, but rather in verifying the accuracy of the free atom model to represent theoretical results over a large temperature range. There are two primary objectives of this report: (1) Verify agreement between MERCURY and TART results, both using continuous energy cross sections. In particular we want to verify the free atom scattering treatment in MERCURY as used over an extended temperature range; by comparison to many other codes for TART this has already been verified over many years [4, 7]. (2) Demonstrate that this agreement depends on ...
Date: August 16, 2006
Creator: Cullen, D E; McKinley, S & Hagmann, C
Partner: UNT Libraries Government Documents Department

Generation of a Broad-Group HTGR Library for Use with SCALE

Description: With current and ongoing interest in high temperature gas reactors (HTGRs), the U.S. Nuclear Regulatory Commission (NRC) anticipates the need for nuclear data libraries appropriate for use in applications for modeling, assessing, and analyzing HTGR reactor physics and operating behavior. The objective of this work was to develop a broad-group library suitable for production analyses with SCALE for HTGR applications. Several interim libraries were generated from SCALE fine-group 238- and 999-group libraries, and the final broad-group library was created from Evaluated Nuclear Data File/B Version ENDF/B-VII Release 0 cross-section evaluations using new ORNL methodologies with AMPX, SCALE, and other codes. Furthermore, intermediate resonance (IR) methods were applied to the HTGR broadgroup library, and lambda factors and f-factors were incorporated into the library s nuclear data files. A new version of the SCALE BONAMI module named BONAMI-IR was developed to process the IR data in the new library and, thus, eliminate the need for the CENTRM/PMC modules for resonance selfshielding. This report documents the development of the HTGR broad-group nuclear data library and the results of test and benchmark calculations using the new library with SCALE. The 81-group library is shown to model HTGR cases with similar accuracy to the SCALE 238-group library but with significantly faster computational times due to the reduced number of energy groups and the use of BONAMI-IR instead of BONAMI/CENTRM/PMC for resonance self-shielding calculations.
Date: June 1, 2012
Creator: Ellis, Ronald James; Lee, Deokjung; Wiarda, Dorothea; Williams, Mark L & Mertyurek, Ugur
Partner: UNT Libraries Government Documents Department

Nuclear data adjustment methodology utilizing resonance parameter sensitivities and uncertainties

Description: This work presents the development and demonstration of a Nuclear Data Adjustment Method that allows inclusion of both energy and spatial self-shielding into the adjustment procedure. The resulting adjustments are for the basic parameters (i.e., resonance parameters) in the resonance regions and for the group cross sections elsewhere. The majority of this development effort concerns the production of resonance parameter sensitivity information which allows the linkage between the responses of interest and the basic parameters. The resonance parameter sensitivity methodology developed herein usually provides accurate results when compared to direct recalculations using existing and well-known cross section processing codes. However, it has been shown in several cases that self-shielded cross sections can be very non-linear functions of the basic parameters. For this reason caution must be used in any study which assumes that a linear relationship exists between a given self-shielded group cross section and its corresponding basic data parameters.
Date: January 1, 1984
Creator: Broadhead, B.L.
Partner: UNT Libraries Government Documents Department

Criticality safety of low-density storage arrays

Description: This note proposes a straightforward and simple method for the criticality safety analysis of fissionable materials configured into large arrays of standard containers. While criticality-safe storage limits have been well-established for standard containers--even under flooded conditions, it is also necessary to rule out the potential for criticality arising from neutronic interactions among multiple containers that might build up over long distances in a large array. Traditionally, the array problem has been approached by individual Monte Carlo analyses of explicit arrangements of single units and their surroundings. Here, the authors show how multiple Monte Carlo analyses can be usefully combined for wide-ranging general application. The technique takes advantage of low average density of fissionable material in typical storage arrays to separate neutron interactions that take place in the neutron`s ``birth unit`` from subsequent interactions in a highly dilute array. Effects of array size, in particular, are conservatively calculated by straightforward analyses which simply smear array contents uniformly across the extent of the array. For given unit loadings in standard containers, practical expressions for neutron multiplication depend only on overall array shape, size and reflective boundary.
Date: May 1, 1996
Creator: Bauer, T.H.
Partner: UNT Libraries Government Documents Department

An ``exact`` treatment of self-shielding and covers in neutron spectra determinations

Description: Most neutron spectrum determination methodologies ignore self-shielding effects in dosimetry foils and treat covers with an exponential attenuation model. This work provides a quantitative analysis of the approximations in this approach. It also provides a methodology for improving the fidelity of the treatment of the dosimetry sensor response to a level consistent with the user`s spectrum characterization approach. A library of correction functions for the energy-dependent sensor response has been compiled that addresses dosimetry foils/configurations in use at the Sandia National Laboratories Radiation Metrology Laboratory.
Date: June 1, 1995
Creator: Griffin, P.J. & Kelly, J.G.
Partner: UNT Libraries Government Documents Department

MERCURY vs. TART Comparisons to Verify Thermal Scattering

Description: Recently the results from many Monte Carlo codes were compared for a series of theoretical pin-cells; the results are documented in ref. [3]; details are also provided here in Appendix A and B. The purpose of this earlier code comparison was primarily to determine how accurately our codes model both bound and free atom neutron thermal scattering. Prior to this study many people assumed that our Monte Carlo transport codes were all now so accurate that they would all produce more or less the same answers, say for example K-eff to within 0.1%. The results demonstrated that in reality we see a rather large spread in the results for even simple scalar parameters, such as K-eff, where we found differences in excess of 2%, far exceeding many people's expectations. The differences between code results were traced to four major factors, (1) Differences between the sets of nuclear data used. (2) The accuracy of nuclear data processing codes. (3) The accuracy of the models used in our Monte Carlo transport codes. (4) Code user selected input options. Naturally at Livermore we would like to insure that we minimize the effects of these factors. In this report we compare the results using two of our Monte Carlo transport codes: MERCURY [2] and TART [2], with the following constraints designed to address the four points listed above, (1) Both codes used exactly the same nuclear data, namely the TART 2005 data. (2) Each code used its own nuclear data processing code. Even though these two data processing codes are independent, they have been extensively tested to insure the processed output results closely agree. (3) Both used the same nuclear physics models. This required that some physics be turned off in each code, namely, (a) Unresolved resonance energy region self-shielding was turned off in ...
Date: March 30, 2006
Creator: Cullen, D E; McKinley, S & Hagmann, C
Partner: UNT Libraries Government Documents Department


Description: To extend their ability to perform accountability and verification measurements of {sup 235}U in a U-Pu oxide matrix, the K-Area Material Storage facility commissioned the development and construction of a Passive/Active {sup 252}Cf Shuffler. A series of {sup 252}Cf, PuO{sub 2}, and U-Pu oxide standards, in addition to a single U{sub 3}O{sub 8} standard, were measured to characterize and calibrate the shuffler. Accompanying these measurements were simulations using MCNP5/MCNPX, aimed at isolating the neutron countrate contributions for each of the isotopes present. Two calibration methods for determining the {sup 235}U content in mixed UPu oxide were then developed, yielding comparable results. The first determines the {sup 235}U mass by estimating the {sup 239}Pu/{sup 235}U ratio-dependent contributions from the primary delayed neutron contributors. The second defines an average linear response based on the {sup 235}U and {sup 239}Pu mass contents. In each case, it was observed that self-shielding due to {sup 235}U mass has a large influence on the observed rates, requiring bounds on the applicable limits of each calibration method.
Date: May 26, 2011
Creator: Dubose, F.
Partner: UNT Libraries Government Documents Department

Continuous Energy, Multi-Dimensional Transport Calculations for Problem Dependent Resonance Self-Shielding

Description: The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system.
Date: March 31, 2009
Creator: Downar, T.
Partner: UNT Libraries Government Documents Department

NJOY 99/2001: new capabilities in data processing

Description: The NJOY Nuclear Data Processing System is used all over the world to process evaluated nuclear data in the ENDF format into libraries for applications. Over the last few years, a number of new capabilities have been added to the system to provide advanced features for MCNP, MCNPX, and other applications codes. These include probability tables for unresolved range self shielding, capabilities optimized for high-energy libraries (typically to 150 MeV for accelerator applications), options for detailed treatments of incident and outgoing charged particles, and a capability to handle photonuclear reactions. These new features and recent experience using NJOY99 for library production will be discussed, along with possible future work, such as delayed-neutron processing and capabilities to handle the new generation of photo-atomic, electro-atomic, and atomic-relaxation evaluations now becoming available in ENDF format. The latest version of the code, NJOY 2001, uses modern Fortran90 style, modularization, and memory allocation methods. The Evaluated Nuclear Data Files (ENDF) format has become the standard for representing nuclear data throughout the world, being used in the US ENDF/B libraries, the European JEF libraries, the Japanese JENDL libraries, and many others. At the same time, the NJOY Nuclear Data Processing System, which is used to convert evaluated nuclear data in the ENDF format into data libraries for nuclear applications, has become the method of choice throughout the world. The combination of these modern libraries of evaluated nuclear data and NJOY processing has proved very capable for classical applications in reactor analysis, fusion work, shielding, and criticality safety. However, over the last few years, new applications have appeared that require extended evaluated data and new processing techniques. A good example of this is the interest in accelerator-boosted applications, which has led to the need for data to higher energies, such as 150 MeV. New kinds of evaluated ...
Date: January 1, 2002
Creator: MacFarlane, R. E. (Robert E.)
Partner: UNT Libraries Government Documents Department

Computation of analytical bounds for cross section self-shielding factors

Description: The shielding factor method (SFM) is a frequently used economical procedure for computing the effective multigroup cross sections needed in reactor analysis. While initially developed and employed in codes used by the fast reactor community, the method has been receiving increased attention in recent years from the electric utility industry, for applications to power reactors. A fundamental problem regarding the method's applicability is to determine the limits of the range of values within which a cross section shielding factor is restricted, and whether these limits are physically meaningful. In a previous paper strict upper and lower bounds for the transport f-factor and for the sum of reaction f-factors were derived and discussed. The purpose of the present work is to present extensions of the methodology used for the upper and lower bounds of the transport f-factor and reaction f-factors.
Date: January 1, 1979
Creator: Barhen, J. & Cacuci, D.G.
Partner: UNT Libraries Government Documents Department

Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks

Description: Versions of MCNP up through and including 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. This paper presents MCNP results for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment.
Date: December 31, 1998
Creator: Mosteller, R.D. & Little, R.C.
Partner: UNT Libraries Government Documents Department

Lump correction and identification in the combined thermal/epithermal neutron (CTEN) method

Description: The authors present a model for self shielding in lumps of fissile material in active-neutron assays. The model combines the formula for self-attenuation of gamma-ray in lumpy sources with the multi-group analysis techniques used in neutron transport calculations. Models for thin foils and for spheres are examined in terms of error multiplication in determining lump corrections and the basic accuracy of the model.
Date: December 31, 1998
Creator: Estep, R.J.; Coop, K.L.; Hollas, C.; Melton, S. & Miko, D.
Partner: UNT Libraries Government Documents Department

SWANS: A Prototypic SCALE Criticality Sequence for Automated Optimization Using the SWAN Methodology

Description: SWANS is a new prototypic analysis sequence that provides an intelligent, semi-automatic search for the maximum k{sub eff} of a given amount of specified fissile material, or of the minimum critical mass. It combines the optimization strategy of the SWAN code with the composition-dependent resonance self-shielded cross sections of the SCALE package. For a given system composition arrived at during the iterative optimization process, the value of k{sub eff} is as accurate and reliable as obtained using the CSAS1X Sequence of SCALE-4.4. This report describes how SWAN is integrated within the SCALE system to form the new prototypic optimization sequence, describes the optimization procedure, provides a user guide for SWANS, and illustrates its application to five different types of problems. In addition, the report illustrates that resonance self-shielding might have a significant effect on the maximum k{sub eff} value a given fissile material mass can have.
Date: January 11, 2001
Creator: Greenspan, E.
Partner: UNT Libraries Government Documents Department

Average Total Neutron Cross Section OF 233U, 235U AND 239Pu from ORELA Transmission Measurements and Statistical Analysis of the Data

Description: The average total neutron cross sections of {sup 233}U, {sup 235}U, and {sup 239}Pu were obtained from transmission measurements in the unresolved resonance region up to several hundred keV neutron energy. The method used for the calculation of the self-shielding effect is described. A statistical model analysis of the results was performed and the s-, p- and d-wave neutron strength functions were obtained.
Date: August 24, 2001
Creator: Derrien, H.
Partner: UNT Libraries Government Documents Department


Description: The objective of this analysis is to develop a repository layout, for Feature No. 13, that will accommodate self-shielding waste packages (WP) with an areal mass loading of 25 metric tons of uranium per acre (MTU/acre). The scope of this analysis includes determination of the number of emplacement drifts, amount of emplacement drift excavation required, and a preliminary layout for illustrative purposes.
Date: April 9, 1999
Creator: Owen, J.
Partner: UNT Libraries Government Documents Department

Average Neutron Total Cross Sections in the Unresolved Energy Range From ORELA High Resolutio Transmission Measurements

Description: Average values of the neutron total cross sections of {sup 233}U, {sup 235}U, {sup 238}U, and {sup 239}Pu have been obtained in the unresolved resonance energy range from high-resolution transmission measurements performed at ORELA in the past two decades. The cross sections were generated by correcting the effective total cross sections for the self-shielding effects due to the resonance structure of the data. The self-shielding factors were found by calculating the effective and true cross sections with the computer code SAMMY for the same Doppler and resolution conditions as for the transmission measurements, using an appropriate set of resonance parameters. Our results are compared to results of previous measurements and to the current ENDF/B-VI data.
Date: May 27, 2004
Creator: Derrien, H
Partner: UNT Libraries Government Documents Department

Calculation of probability table parameters to include intermediate resonance self-shielding

Description: In order to demonstrate the practicality of the multi-band method (Cullen, Nucl. Sc. Eng., 55, 387 (1974)) as applied to all energy ranges, it is demonstrated that intermediate resonance effects may be included; usually only two or three, and at most four, bands are required in any cross section probability table; and a low-order rational approximation is an excellent means of defining Bonderenko sigma/sub 0/ self-shielded cross sections. 1 table.
Date: July 1, 1977
Creator: Cullen, D.E.
Partner: UNT Libraries Government Documents Department

Background cross section method as a general tool for reactor analysis

Description: The background cross section method (also called the self-shielding method) has been used extensively in fast reactor analysis. More recently it has also become important in thermal power reactor studies. This paper reviews current applications of the method and describes efforts underway at the Los Alamos Scientific Laboratory to improve the accuracy and reliability of the approach and to extend its applicability to graphite moderated systems and shielding problems. Improvements discussed include a method for automatically accounting for energy dependent buckling that resolves long-standing discrepancies in the calculation of iron reflected criticals and which promises to improve deep-penetration calculations in iron, methods for treating intermediate resonance effects, methods for treating double heterogeneity in gas-cooled reactors, the automatic calculation of Levine factors, improved treatments of elastic removal, and improvements in processing codes and formats.
Date: January 1, 1978
Creator: MacFarlane, R.E.; Kidman, R.B.; LaBauve, R.J. & Becker, M.
Partner: UNT Libraries Government Documents Department

Measurements of isotope effects in the photoionization of N2 and implications for Titan's atmosphere

Description: Isotope effects in the non-dissociative photoionization of molecular nitrogen (N2 + h nu -> N2+ + e-) may play a role in determining the relative abundances of isotopic species containing nitrogen in interstellar clouds and planetary atmospheres but have not been previously measured. Measurements of the photoionization efficiency spectra of 14N2, 15N14N, and 15N2 from 15.5 to 18.9 eV (65.6-80.0 nm) using the Advanced Light Source at Lawrence Berkeley National Laboratory show large differences in peak energies and intensities, with the ratio of the energy-dependent photoionization cross-sections, sigma(14N2)/sigma(15N14N), ranging from 0.4 to 3.5. Convolving the cross-sections with the solar flux and integrating over the energies measured, the ratios of photoionization rate coefficients are J(15N14N)/J(14N2)=1.00+-0.02 and J(15N2)/J(14N2)=1.00+-0.02, suggesting that isotopic fractionation between N2 and N2+ should be small under such conditions. In contrast, in a one-dimensional model of Titan's atmosphere, isotopic self-shielding of 14N2 leads to values of J(15N14N)/J(14N2) as large as ~;;1.17, larger than under optically thin conditions but still much smaller than values as high as ~;;29 predicted for N2 photodissociation. Since modeled photodissociation isotope effects overpredict the HC15N/HC14N ratio in Titan's atmosphere, and since both N atoms and N2+ ions may ultimately lead to the formation of HCN, estimates of the potential of including N2 photoionization to contribute to a more quantitative explanation of 15N/14N for HCN in Titan's atmosphere are explored.
Date: December 30, 2010
Creator: Croteau, Philip; Randazzo, John B.; Kostko, Oleg; Ahmed, Musahid; Liang, Mao-Chang; Yung, Yuk L. et al.
Partner: UNT Libraries Government Documents Department