Oak Ridge National Laboratory
Post Office Box 2008 Oak Ridge, Tennessee 37831-6362
managed by Lockheed Martin Energy Research Corp.
for the U.S. Department of Energy
Phone No. 423-574-6176
FAX 423-574-6182
Internet: PDC@ORNL.GOV
WWW: http://epicws.epm.ornl.gov
No. 383 October 1996
A new version of the SCALE Training Course was offered for the first time at ORNL. (SCALE is a modular code system used by approximately 800 persons worldwide for criticality safety, shielding, and heat transfer calculations of nuclear facility and package designs.) Based on suggestions from attendees of previous SCALE courses, the basic week-long course was expanded to two four-day segments. The first segment of the course was held October 15-18, 1996, and included 25 attendees. This segment covered the criticality analysis sequences within the SCALE system. The second segment of the course was held October 21-24, 1996, and included 21 attendees. This segment covered source characterizations and shielding sequences in SCALE. Nine attendees participated in both segments of the course; four participatns were from Asia and Europe. Within the U.S., there was representation from nuclear utilities, consultants, DOE, and the NRC. Based on evaluations by the attendees, the new format was successful.
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| CCC-610/CALOR95 OP SYS: UNIX Language: Fortran 77 Computers: Sun/IBM workstation Format: tar | Oak Ridge National Laboratory contributed a new software version of
this Monte Carlo code system for design and analysis of calorimeter
systems, Spallation Neutron Source (SNS) target systems, etc. The
new version runs on Sun Sparc workstations under Solaris. CALOR95
was designed to assist experimentalists in evaluating and analyzing
different types of calorimeter systems used in many high-energy
physics experiments to determine the energy and direction of incident
hadrons, leptons, and photons. This code package contains HETC95,
SPECT95, EGS4, MORSE, and other support programs. CALOR95 is available for either IBM RS/6000 or Sun workstations and requires a Fortran 77 compiler. It is transmitted on either one CD-ROM, DC 6150 (150 MB), 4-mm DAT (8 GB), or 8-mm (2.3 GB) cartridge tape in compressed Unix tar format. References: ORNL/TM-11185 (unpublished report) and SDC-92-00257 (May 1992). Fortran 77; IBM RS/6000 (C00610/IRISC/01) or Sun Sparc (C00610/SUN05/00). PSR-137/MARLOWE
14 | OP SYS: many Language: Fortran 77, C Computers: many Format: DOS, tar Oak Ridge National Laboratory contributed a newly frozen version of
this code system for simulating atomic collisions in crystalline targets
using the binary collision approximation. This new release is
designated Version 14 and includes many enhancements and some
corrections to the scripts, preprocessor, control files, and source files.
All of the changes are too numerous to list here, but these are just a
few. The management of the keyword dictionary was improved and
support for TABULA was added. The algorithms for finding the apsis
in a collision and for evaluating the 'time' integral were replaced,
allowing MARLOWE potentials to have an attractive region. The new
time integral formulation is more accurate and the results of the test
problems are altered in several respects. A sixth interatomic potential
energy function is supplied. The scale identifying the status of cascade
atoms was revised and expanded to make space for additional features.
An analysis of the final states of focusons and replacement sequences
was added. The histogram management procedure BOXER was
redesigned as a set of less interwoven procedures. | The system executes on IBM MVS, VAX VMS, Data General, MS-DOS and a variety of Unix systems (AIX, DG/UX, EP/IX, HP-UX, IRIX, DEC Unix, SunOS, Ultrix, and Unicos.) MARLOWE runs on personal computers under Microsoft DOS, version 6 using the Lahey Fortran 90 compiler Version 1.10h. Source, scripts, test problems, and documentation files are included in the package. No executables are included in this package, which is transmitted either on one DS/HD 3.5-in. (1.44 MB) diskette in self-extracting compressed DOS files or on one diskette in a compressed tar file. References: Unpublished report (1996), Phys. Rev. B 40, (1989), Nucl. Instrum. Methods in Phys. Res. B 48 (1990), and Nucl. Instrum. Methods in Phys. Res. B 67 (1992). Fortran 77 and C or Assembler; many computers (P00137/MNYCP/02). PSR-363/FRANCO | OP SYS: many Language: WATFOR77 Computers: many Format: DOS Penn State University contributed a new finite element fuel rod
analysis code system called FRANCO, which is a quasi-static two-dimensional code that calculates the fuel temperature and material
deformation as a function of heat generation rate. Both solid and
annular fuel configurations are modeled. FRANCO uses finite element
theory and applications for mechanical deformation and heat
conduction and determines the temperature distribution from the fuel
center to the coolant adjacent to the clad at a position along the fuel
rod axis. It calculates the average temperature of each radial division,
the nodal displacement, and strain and stress within the fuel pellet and
clad. The principal stresses which represent maximum and minimum
stresses within an element result from Mohr's circle relationship
between normal stresses. FRANCO is capable of predicting the
thermo-mechanical behavior in the radial direction of a single fuel rod
for both boiling water reactors (BWR's) and pressurized water
reactors (PWR's). The cross sectional plane geometry of the fuel rod
is modeled using three-node constant strain triangular finite elements. | A time-dependent problem can be simulated using quasi-static analysis when time-dependent parameters are provided. FRANCO can treat a steady-state or mild transient condition as a time-independent problem, where time-dependent parameters are input as "snap shots." Consequently, the FRANCO code is able to run faster and easier compared to other industrial codes with comparable results; and hence, FRANCO is useful in analyzing the fuel element performance as a scoping tool. FRANCO runs on IBM personal computers and mainframes. It is written in WATFOR77, and the WATCOM compiler was used to build the IBM PC executables included in the package. At RSICC the packaged executables were tested on an IBM PC 486/DX2 66 and on an IBM Pentium under MS-DOS as bundled with Windows 95. FRANCO was also tested using the Lahey LF90 compiler (Version 1.10h) under MS-DOS as bundled with Windows 95 on an IBM PC 486. The Fortran sources were also compiled using the GNU F77 compiler Version 2.7.2 under Slackware Linux 2.0 on an IBM Pentium (60 MHZ). The package is transmitted on one 1.44 DS/HD 3.5-in. diskette written in self-extracting compressed DOS files which include the FRANCO source, PC executables, test cases and documentation files. References: Penn State University Master Thesis (December 1994). |
| DLC-154/ANSLV OP SYS: many Language: NA Computers: all Format: tar | Oak Ridge National Laboratory, Oak Ridge, Tennessee contributed a
newly frozen version of these pseudo-problem-independent,
multigroup cross section libraries, which were generated to support
the Advanced Neutron Source (ANS) reactor design studies. There
were many changes in this new release which includes additional
nuclides, revisions of several nuclides, and the removal of some
nuclides for a current total of 138 isotopes. Details of these changes
are included in a new supplemental report which is distributed with
RSICC's documentation. The ANS was a proposed reactor which
would have been fueled with highly enriched uranium and cooled with
heavy water. The libraries, designated ANSL-V (Advanced Neutron
Source Cross Section Libraries based on ENDF/B-V) are data based
in AMPX master format. ANSL-V data can be used for the
subsequent generation of problem-dependent fine- and/or broad-group
cross sections for a wide range of applications, such as core and shield
analysis, activation analyses after irradiation of certain elements in the
reactor environment, and safety analyses. Problem-dependent cross
sections can be derived from the ANSL-V data with the AMPX
modular code system (PSR-315/AMPX-77). ANSL-V consists of the following fine and broad groups:
The AIM module of PSR-315/AMPX-77, PSR-352/SCAMPI, or CCC-545/SCALE 4.3 can be used for mode conversion of the data. Some of the AMPX utility modules are included in SCAMPI, and SCALE 4.3 and may also be used with the libraries. Note that previous versions of AMPX and SCALE will not work because of the AMPX master library format changes. References: ORNL-6618/s1 (August 1995) and ORNL-6618 (September 1990). D00154/ALLCP/01. |
In serving a specialized area of scientific endeavor, it seems important that we note significant events or changes in the activities of people concerned with radiation protection, transport, and shielding in the nuclear industry. We, therefore, continue to carry personal items as they are brought to our attention.
Dr. Lee L. Carter formerly of the Westinghouse Hanford Company, has formed his own consulting company, Carter M.C. Analysis, P.O. Box 234, Richland, WA 99352-0234 (phone 509-627-1301, email Glcar39@aol.com).
The following literature cited has been ordered for review, and that selected as suitable will be placed in the RSICC Information Storage and Retrieval Information System (SARIS). This early announcement is made as a service to the shielding community. Copies of the literature are not distributed by RSICC. They may generally be obtained from the author or from a documentation center such as the National Technical Information Service (NTIS), Department of Commerce, Springfield, Virginia 22161. For literature listed as available from INIS contact INIS Clearinghouse, International Atomic Energy Agency, P.O. Box 100, A-1400 Vienna.
RSICC maintains a microfiche file of the literature entered into SARIS, and duplicate copies of out-of-print reports may be available on request. Naturally, we cannot fill requests for literature which is copyrighted (such as books or journal articles) or whose distribution is restricted.
This literature is on order. It is not in our system. Please order from NTIS or other available source as indicated.
Health Phys.., 71, 425-437 . . . Overview of the Fernald Dosimetry Reconstruction Project and Source Term Estimates For 1951-1988. . . . Meyer, K.R.; Voileque, P.G.; Schmidt, D.W.; Rope, S.K.; Killough, G.G.; Shleien, B.; Moore, . . . October 1996 . . . Keystone Scientific, Fort Collins, CO; MJP Risk Keystone Scientific, Fort Collins, CO; MJP Risk Assessment, Inc., Idaho Falls, ID; Health Physics Applications, Darnestown, MD; Environmental Perspectives, Inc., Idaho Falls, ID; Hendecagon Corp., Oak Ridge, TN; Scinta, Inc., Silver Spring, MD; Moore Technical Associates, Inc., Oak Ridge, TN; Eagle Rock Scientific, Idaho Falls, ID; Radiological Assessments Corp., Neeses, SC.
Health Phys., 71, 438-456 . . . Dose Assessment Activities in the Republic of the Marshall Islands. . . . Simon, S.L.; Graham, J.C. . . . October 1996 . . . Nationwide Radiological Study, Majuro, Marshall Islands.
Health Phys., 71, 457-469 . . . Identification and Screening Evaluation of Key Historical Materials and Emission Sources at the Oak Ridge Reservation. . . . Widner, T.E.; Ripple, S.R.; Buddenbaum, J.E. . . . October 1996 . . . McLaren/Hart Environmental Services, Inc., Alameda, CA and Cleveland, OH.
Health Phys., 71, 470-476 . . . Challenges in Developing Estimates of Exposure Rate Near the Nevada Test Site. . . . Thompson, C.B.; McArthur, R.D. . . . October 1996 . . . University and Community College System of Nevada, Las Vegas, NV.
Health Phys., 71, 477-486 . . . Ingestion of Nevada Test Site Fallout: Internal Dose Estimates. . . . Whicker, F.W.; Kirchner, T.B.; Anspaugh, L.R.; Ng, Y.C. . . . October 1996 . . . Colorado State University, Fort Collins, CO; Lawrence Livermore National Laboratory, Livermore, CA.
Health Phys., 71, 487-501 . . . Estimating Internal Dose Due to Ingestion of Radionuclides from Nevada Test Site Fallout. . . . Kirchner, T.B.; Whicker, F.W.; Anspaugh, L.R.; Ng, Y.C. . . . October 1996 . . . Colorado State University, Fort Collins, CO; Lawrence Livermore National Laboratory, Livermore, CA.
Health Phys., 71, 502-509 . . . Past Radionuclide Releases from Routine Operations at Rocky Flats. . . . Ripple, S.R.; Widner, T.E.; Mongan, T.R. . . . October 1996 . . . McLaren/Hart Environmental Engineering, Alameda, CA .
Health Phys., 71, 510-521 . . . Plutonium Releases from the 1957 Fire at Rocky Flats. . . . Mongan, T.R.; Ripple, S.R.; Brorby, G.P.; diTommaso, D.G. . . . October 1996 . . . McLaren/Hart Environmental Engineering, Inc., Alameda, CA; Geomatrix Consultants, Inc., San Francisco, CA.
Health Phys., 71, 522-531 . . . Plutonium Release From the 903 Pad at Rocky Flats. . . . Mongan, T.R.; Ripple, S.R.; Winges, K.D. . . . October 1996 . . . McLaren/Hart Environmental Engineering, Inc., Alameda, CA; McCulley, Frick & Gilman, Inc., Lynnwood, WA.
Health Phys., 71, 532-544 . . . Hanford Environmental Dose Reconstruction Project - An Overview. . . . Shipler, D.B.; Napier, B.A.; Farris, W.T.; Freshley, M.D. . . . October 1996 . . . Battelle, Pacific Northwest Labs., Richland, WA.
Health Phys., 71, 545-555 . . . Reconstruction of Radionuclide Releases from the Hanford Site, 1944-1972. . . . Heeb, C.M.; Gydesen, S.P.; Simpson, J.C.; Bates, D.J. . . . October 1996 . . . Battelle, Pacific Northwest Labs., Richland, WA.
Health Phys., 71, 556-567 . . . Reconstruction of Radioactive Contamination in the Columbia River. . . . Walters, W.H.; Richmond, M.C.; Gilmore, B.G. . . . October 1996 . . . (retired) Kennewick, WA;, Battelle, Pacific Northwest Labs., Richland, WA.
Health Phys., 71, 568-577 . . . Atmospheric Dispersion and Deposition of 131I Released from the Hanford Site. . . . Ramsdell, J.V.; Simonen, C.A.; Burk, K.W.; Stage, S.A. . . . October 1996 . . . Battelle, Pacific Northwest Labs., Richland, WA.
Health Phys., 71, 578-587 . . . Developing Historical Food Production and Consumption Data for 131I Dose Estimates: The Hanford Experience. . . . Anderson, D.M.; Marsh, T.L.; Deonigi, D.A. . . . October 1996 . . . Battelle, Pacific Northwest Labs., Richland, WA.
Health Phys., 71, 588-601 . . . Radiation Doses from Hanford Site Releases to the Atmosphere and the Columbia River. . . . Farris, W.T.; Napier, B.A.; Ikenberry, T.A.; Shipler, D.B. . . . October 1996 . . . Battelle, Pacific Northwest Labs., Richland, WA; GRAM, Inc., Los Alamos, NM.
Nucl. Sci. Eng., 123, 1-16 . . . Calculation and Evaluation of Cross Sections and Kerma Factors for Neutrons up to 100 MeV on 16O and 14N. . . . Chadwick, M.B.; Young, P.G. . . . May 1996 . . . Lawrence Livermore National Laboratory, Livermore, CA; Los Alamos National Laboratory, Los Alamos, NM.
Nucl. Sci. Eng., 123, 17-37 . . . Calculation and Evaluation of Cross Sections and Kerma Factors for Neutrons up to 100 MeV on Carbon. . . . Chadwick, M.B.; Cox, L.J.; Young, P.G.; Meigooni, A.S. . . . . . . Lawrence Livermore National Laboratory, Livermore, CA; Los Alamos National Laboratory, Los Alamos, NM; University of Kentucky, Lexington, KY.
Nucl. Sci. Eng., 123, 38-56 . . . The Even-Parity and Simplified Even-Parity Transport Equations in Two-Dimensional x-y Geometry. . . . Taewan Noh; Miller, W.F., Jr.; Morel, J.E. . . . May 1996 . . . University of California, Berkeley, Berkeley, CA; Los Alamos National Laboratory, Los Alamos, NM.
Nucl. Sci. Eng., 123, 57-67 . . . A Locally Exact Numerical Scheme with Nonoscillation Properties for Stationary Transport Equations with Absorption and Source Terms. . . . Sakai, K. . . . May 1996 . . . Saitama Institute of Technology, Saitama, Japan.
Nucl. Sci. Eng., 123, 68-85 . . . Singular Perturbation Solutions of the Neutron Transport Equation. . . . Losey, D.C.; Lee, J.C.; Martin, W.R.; Adamson, T.C. . . . May 1996 . . . Westinghouse Savannah River Co., Aiken, SC; University of Michigan, Ann Arbor, MI.
Nucl. Sci. Eng., 123, 96-109 . . . Continuous Energy Monte Carlo Calculations of Randomly Distributed Spherical Fuels in High-Temperature Gas-Cooled Reactors Based on a Statistical Geometry Model . . . Murata, I.; Mori, T.; Nakagawa, N. . . . May 1996 . . . Japan Atomic Energy Research Institute, Ibaraki-ken, Japan.
Nucl. Sci. Eng., 123, 110-120 . . . The Two-Region Milne Problem. . . . Ganapol, B.D.; Pomraning, G.C. . . . May 1996 . . . NASA/AMES Research Center, Moffett Field, CA; University of California, Los Angeles, CA.
Nucl. Sci. Eng., 123, 121-135 . . . The Effective Fuel Temperature to be Used for Calculating Resonance Absorption in a 238UO2 Lump with a Nonuniform Temperature Profile. . . . de Kruijf, W.J.M.; Janssen, A.J. . . . May 1996 . . . Netherlands Energy Research Foundation ECN, Petten, The Netherlands.
Nucl. Sci. Eng., 124, 228-242 . . . Transmission Through Shields of Quasi-Monoenergetic Neutrons Generated by 43- and 68-MeV Protons - I: Concrete Shielding Experiment and Calculation for Practical Application . . . Nakao, N.; Nakashima, H.; Nakamura, T.; Tanaka, S-I.; Tanaka, S.; Shin, K.; Baba, M.; Sakamo . . . October 1996 . . . Tohoku University, Sendai, Japan; Japan Atomic Energy Research Institute, Japan; Kyoto University, Kyoto, Japan.
Nucl Sci. Eng., 124, 243-257 . . . Transmission Through Shields of Quasi-Monoenergetic Neutrons Generated by 43- and 68-MeV Protons - II: Iron Shielding Experiment and Analysis for Investigating Calculational Method and Cross-Section Data. . . . Nakashima, H.; Nakao, N.; Tanaka, S-I.; Nakamura, T.; Shin, K.; Tanaka, S.; Takada, H.; Meig . . . October 1996 . . . Japan Atomic Energy Research Institute, Ibaraki, Japan; Tohoku University, Sendai, Japan; Kyoto University, Kyoto, Japan.
Nucl. Sci. Eng., 124, 258-270 . . . Effects of the Photon Cross Sections and Energy-Absorption Coefficients of Air to the Gamma-Ray Point Isotropic Exposure Buildup Factors. . . . Hirayama, H. . . . October 1996 . . . National Laboratory for High Energy Physics, Ibaraki, Japan.
Nucl. Sci. Eng., 124, 280-290 . . . Analysis of Nuclear Power Transmutation Potential at Equilibrium. . . . Salvatores, M.; Slessarev, I.; Tchistiakov, A. . . . October 1996 . . . Commissariat a l'Energie Atomique, Cadarache, France.
Nucl. Sci. Eng., 124, 291-308 . . . Optimization of Nonanalog Monte Carlo Games Using Differential Operator Sampling. . . . Sarkar, P.K.; Rief, H. . . . October 1996 . . . Commission of the European Communities, Ispra, Italy.
Nucl. Sci. Eng., 124, 309-319 . . . A Modified P3-Like Approximation to the Transport Equation. . . . Bingjing Su; Pomraning, G.C. . . . October 1996 . . . University of California, Los Angeles, CA.
Nucl. Sci. Eng., 124, 320-326 . . . Boundary Condition Perturbations in Transport Theory. . . . Rahnema, F. . . . October 1996 . . . Georgia Institute of Technology, Atlanta, GA.
Nucl. Sci. Eng., 124, 327-338 . . . Quantitative Method to Measure Void Fraction of Two-Phase Flow Using Electronic Imaging with Neutrons. . . . Mishima, K.; Hibiki, T. . . . October 1996 . . . Kyoto University, Osaka, Japan.
Nucl. Sci. Eng., 124, 349-357 . . . A Semiempirical Method to Calculate Continuum Gamma-Ray Spectra and Multiplicities from Neutron-Induced Reactions. . . . Sheng Fan; Zhixiang Zhao . . . October 1996 . . . Chinese Nuclear Data Center, Beijing, China.
Nucl. Sci. Eng., 124, 358-360 . . . A Note on the Pn Method with Mark Boundary Conditions. . . . Garcia, R.D.M.; Siewert, C.E. . . . October 1996 . . . Centro Tecnico Aeroespacial, Sao Jose dos Campos, Brazil; North Carolina State University, Raleigh, NC.
Nucl. Sci. Eng., 124, 361-363 . . . Autoregressive Moving Average Inverse Dynamics for Rhodium Self-Powered Neutron Detector. . . . Mayo, C.W.; Yan. L. . . . October 1996 . . . North Carolina State University, Raleigh, NC.
Nucl. Sci. Eng., 124, 364-367 . . . Letter On "Neutron Fluence at the Pressure Vessel of a Pressurized Water Reactor Determined by the MCNP Code." . . . Haghighat, A. . . . October 1996 . . . Pennsylvania State University, University Park, PA.
Nucl. Sci. Eng., 124, 367-368 . . . Reply to "On 'Neutron Fluence at the Pressure Vessel of a Pressurized Water Reactor Determined by the MCNP Code.'" . . . Laky, P. Tsoulfanidis, N. . . . October 1996 . . . University of Missouri - Rolla, Rolla, MO.
Nucl. Technol., 116, 1-8 . . . Pellet Power Ratio in a Pellet-Suspension Fission Core. . . . Kingdon, D.R.; Harms, A.A. . . . October 1996 . . . McMaster University, Ontario, Canada.
Nucl. Technol., 116, 9-18 . . . Study on Reactivity Worth of Beryllium By (n,2n) and (y,n) Reactions. . . . Misawa, T.; Shiroya, S.; Kanda, K. . . . October 1996 . . . Nagoya University, Nagoya, Japan; Kyoto University, Osaka, Japan.
Nucl. Technol., 116, 34-54 . . . A Model for Nonvolatile Fission Product Release During Reactor Accident Conditions. . . . Lewis, B.J.; Andre, B.; Ducros, G.; Maro, D. . . . October 1996 . . . Commissariat a l'Energie Atomique, France.
Nucl. Technol., 116, 78-90 . . . Boron-Nitride-Coated Nuclear Fuels. . . . Gunduz, G.; Uslu, I.; Durmazucar, H.H. . . . October 1996 . . . Orta Dogu Teknik Universitesi, Ankara, Turkey; Turkiye Atom Enerjisi Kurumu, Ankara, Turkey; Cumhuriyet Universitesi, Sivas, Turkey.
Nucl. Technol., 116, 91-107 . . . Improvements in Boiling Water Reactor Uranium Utilization and Operating Experience with Burnup Increase. . . . Mochida, T.; Haikawa, K.; Yamashita, J-I.; Nishimura, A.; Iwata, Y.; Arai, S. . . . October 1996 . . . Hitachi Works of Hitachi, Ibaraki, Japan; Tokyo Electric Power Co., Tokyo, Japan.
HPS N13.22-1995 . . . An American National Standard - Bioassay Programs For Uranium. . . . . . . October 1995 . . . Health Physics Society, McLean VA.
HPS N13.32-1995 . . . American National Standard - Performance Testing of Extremity Dosimeters.. . . August 1995 . . . Health Physics Society, McLean, VA.
HPS N13.30-1996 . . . An American National Standard - Performance Criteria for Radiobioassay
. . . May 1996 . . .Health Physics Society, McLean, VA.
INDC(NS)-353 . . . Improvement of Measurements, Theoretical Computations and Evaluations of Neutron Induced Helium Production Cross Sections. . . . Pashchenko, A.B., ed. . . . September 1996 . . . IAEA Nuclear Data Section, Vienna, Austria.
LA-SUB-95-224 . . . TWODANT Benchmark. Progress Report. . . . Lee, S. . . . January 1994 . . . Arizona University, Tucson, AZ.
ORNL/TM-12294/V5 . . . Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 5 - North Anna Unit 1 Cycle 5. . . . Bowman, S.M. . . . October 1996 . . . Oak Ridge National Laboratory, Oak Ridge, TN.
ORNL/TM-13317 . . . An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel. . . . DeHart, M.D.; Hermann, O.W. . . . September 1996 . . . Oak Ridge National Laboratory, Oak Ridge, TN.