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Calculation of cell volumes and surface areas in MCNP

Description: MCNP is a general Monte Carlo neutron-photon particle transport code which treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces. It is necessary to calculate cell volumes and surface areas so that cell masses, fluxes, and other important information can be determined. The volume/area calculation in MCNP computes cell volumes and surface areas for cells and surfaces rotationally symmetric about any arbitrary axis. 5 figures, 1 table.
Date: January 1, 1980
Creator: Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

Random number stride in Monte Carlo calculations

Description: Monte Carlo radiation transport codes use a sequence of pseudorandom numbers to sample from probability distributions. A common practice is to start each source particle a predetermined number of random numbers up the pseudorandom number sequence. This number of random numbers skipped between each source particles the random number stride, S. Consequently, the jth source particle always starts with the j{center dot}Sth random number providing correlated sampling'' between similar calculations. A new machine-portable random number generator has been written for the Monte Carlo radiation transport code MCNP providing user's control of the random number stride. First the new MCNP random number generator algorithm will be described and then the effects of varying the stride will be presented. 2 refs., 1 fig.
Date: January 1, 1990
Creator: Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

Finite difference solution of the time dependent neutron group diffusion equations

Description: In this thesis two unrelated topics of reactor physics are examined: the prompt jump approximation and alternating direction checkerboard methods. In the prompt jump approximation it is assumed that the prompt and delayed neutrons in a nuclear reactor may be described mathematically as being instantaneously in equilibrium with each other. This approximation is applied to the spatially dependent neutron diffusion theory reactor kinetics model. Alternating direction checkerboard methods are a family of finite difference alternating direction methods which may be used to solve the multigroup, multidimension, time-dependent neutron diffusion equations. The reactor mesh grid is not swept line by line or point by point as in implicit or explicit alternating direction methods; instead, the reactor mesh grid may be thought of as a checkerboard in which all the ''red squares'' and '' black squares'' are treated successively. Two members of this family of methods, the ADC and NSADC methods, are at least as good as other alternating direction methods. It has been found that the accuracy of implicit and explicit alternating direction methods can be greatly improved by the application of an exponential transformation. This transformation is incompatible with checkerboard methods. Therefore, a new formulation of the exponential transformation has been developed which is compatible with checkerboard methods and at least as good as the former transformation for other alternating direction methods. (auth)
Date: August 1, 1975
Creator: Hendricks, J.S. & Henry, A.F.
Partner: UNT Libraries Government Documents Department

New methods for neutron response calculations with MCNP

Description: MCNP4B was released for international distribution in February, 1997. The author summarized the new MCNP4B features since the release of MCNP4A over three years earlier and compare some results. Then he describes new methods being developed for future code releases. The focus is methods and applications of ex-core neutron response calculations.
Date: May 1, 1997
Creator: Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

MCNPX Model/Table Comparison

Description: MCNPX is a Monte Carlo N-Particle radiation transport code extending the capabilities of MCNP4C. As with MCNP, MCNPX uses nuclear data tables to transport neutrons, photons, and electrons. Unlike MCNP, MCNPX also uses (1) nuclear data tables to transport protons; (2) physics models to transport 30 additional particle types (deuterons, tritons, alphas, pions, muons, etc.); and (3) physics models to transport neutrons and protons when no tabular data are available or when the data are above the energy range (20 to 150 MeV) where the data tables end. MCNPX can mix and match data tables and physics models throughout a problem. For example, MCNPX can model neutron transport in a bismuth germinate (BGO) particle detector by using data tables for bismuth and oxygen and using physics models for germanium. Also, MCNPX can model neutron transport in UO{sub 2}, making the best use of physics models and data tables: below 20 MeV, data tables are used; above 150 MeV, physics models are used; between 20 and 150 MeV, data tables are used for oxygen and models are used for uranium. The mix-and-match capability became available with MCNPX2.5.b (November 2002). For the first time, we present here comparisons that calculate radiation transport in materials with various combinations of data charts and model physics. The physics models are poor at low energies (<150 MeV); thus, data tables should be used when available. Our comparisons demonstrate the importance of the mix-and-match capability and indicate how well physics models work in the absence of data tables.
Date: March 3, 2003
Creator: Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

Computational benchmark for deep penetration in iron

Description: A benchmark for calculation of neutron transport through iron is now available based upon a rigorous Monte Carlo treatment of ENDF/B-IV and ENDF/B-V cross sections. The currents, flux, and dose (from monoenergetic 2, 14, and 40 MeV sources) have been tabulated at various distances through the slab using a standard energy group structure. This tabulation is available in a Los Alamos Scientific Laboratory report. The benchmark is simple to model and should be useful for verifying the adequacy of one-dimensional transport codes and multigroup libraries for iron. This benchmark also provides useful insights regarding neutron penetration through iron and displays differences in fluxes calculated with ENDF/B-IV and ENDF/B-V data bases. (GHT)
Date: October 1, 1979
Creator: Carter, L.L. & Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

/sup 3/He detector design for low-level transuranic waste assay

Description: From this study it appears that a logical configuration of /sup 3/He detectors imbedded in CH/sub 2/ for nondestructive assay of low-level transuranic waste is: rho = 1 atmosphere, d = 2.54 cm, x = 5.08 cm, t = 15.24 cm, h = 61 cm, and N = 5. If a greater detector response is desired, the best way to achieve it is to first double the detector height, h, and then to increase the number of tubes, N, and/or increase the /sup 3/He density, rho.
Date: January 1, 1978
Creator: Hendricks, J.S. & Close, D.A.
Partner: UNT Libraries Government Documents Department

MCNP variance reduction overview

Description: The MCNP code is rich in variance reduction features. Standard variance reduction methods found in most Monte Carlo codes are available as well as a number of methods unique to MCNP. We discuss the variance reduction features presently in MCNP as well as new ones under study for possible inclusion in future versions of the code.
Date: January 1, 1985
Creator: Hendricks, J.S. & Booth, T.E.
Partner: UNT Libraries Government Documents Department

MCNP4A: Features and philosophy

Description: This paper describes MCNP, states its philosophy, introduces a number of new features becoming available with version MCNP4A, and answers a number of questions asked by participants in the workshop. MCNP is a general-purpose three-dimensional neutron, photon and electron transport code. Its philosophy is ``Quality, Value and New Features.`` Quality is exemplified by new software quality assurance practices and a program of benchmarking against experiments. Value includes a strong emphasis on documentation and code portability. New features are the third priority. MCNP4A is now available at Los Alamos. New features in MCNP4A include enhanced statistical analysis, distributed processor multitasking, new photon libraries, ENDF/B-VI capabilities, X-Windows graphics, dynamic memory allocation, expanded criticality output, periodic boundaries, plotting of particle tracks via SABRINA, and many other improvements. 23 refs.
Date: May 1, 1993
Creator: Hendricks, J. S.
Partner: UNT Libraries Government Documents Department

MCNP{trademark} directions

Description: The MCNP code development program is a relatively large and rapidly changing project in the small and highly-specialized field of radiation transport, specifically radiation protection and shielding. A number of major new MCNP initiatives are described in the subsequent papers in this session. The focus of this paper is the important new developments not described elsewhere and a number of recent developments that have been available since MCNP4A but have gone unnoticed. In particular, we report for the first time a new MCNP quality assurance initiative providing 97% test coverage, a new MCNP feature enabling plotting of nuclear data, and the other new features developed so far for MCNP4B. Finally, an attempt is made to articulate how all these fit together into the overall MCNP development program.
Date: August 1, 1994
Creator: Hendricks, J. S.
Partner: UNT Libraries Government Documents Department

MCNPX version 2.5.c

Description: MCNPX is a Fortran 90 Monte Carlo radiation transport computer code that transports all particles at all energies. It is a superset of MCNP4C3, and has many capabilities beyond MCNP4C3. These capabilities are summarized along with their quality guarantee and code availability. Then the user interface changes from MCNP are described. Finally, the n.ew capabilities of the latest version, MCNPX 2.5.c, are documented. Future plans and references are also provided.
Date: January 1, 2003
Creator: Hendricks, J. S. (John S.)
Partner: UNT Libraries Government Documents Department

Performance of scientific computing platforms running MCNP4B

Description: A performance study has been made for the MCNP4B Monte Carlo radiation transport code on a wide variety of scientific computing platforms ranging from personal computers to Cray mainframes. We present the timing study results using MCNP4B and its new test set and libraries. This timing study is unlike other timing studies because of its widespread reproducibility, its direct comparability to the predecessor study in 1993, and its focus upon a nuclear engineering code. Our results, derived from using the new 29-problem test set for MCNP4B, (1) use a highly versatile and readily available physics code; (2) show that timing studies are very problem dependent; (3) present the results as raw data allowing comparisons of performance to other computing platforms not included in this study to those platforms that were included; (4) are reproducible; and (5) provide a measure of improvement in performance with the MCNP code due to advancements in software and hardware over the past 4 years. In the 1993 predecessor study using MCNP4A, the performances were based on a 25 problem test set. We present our data based on MCNP4B`s new 29 problem test set which cover 97% of all the FORTRAN physics code lines in MCNP4B. Like the previous study the new test problems and the test data library are available from the Radiation Shielding and Information Computational Center (RSICC) in Oak Ridge, Tennessee. Our results are reproducible because anyone with the same workstation, compiler, and operating system can duplicate the results presented here. The computing platforms included in this study are the Sun Sparc2, Sun Sparc5, Cray YMP 8/128, HP C180,SGI origin 2000, DBC 3000/600, DBC AiphaStation 500(300 MHz), IBM RS/6000-590, HP /9000-735, Micron Milienia Pro 200 MHz PC, and the Cray T94/128.
Date: November 1, 1997
Creator: McLaughlin, H.E. & Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

An enhanced geometry-independent mesh weight window generator for MCNP

Description: A new, enhanced, weight window generator suite has been developed for MCNP{trademark}. The new generator correctly estimates importances in either an user-specified, geometry-independent orthogonal grid or in MCNP geometric cells. The geometry-independent option alleviates the need to subdivide the MCNP cell geometry for variance reduction purposes. In addition, the new suite corrects several pathologies in the existing MCNP weight window generator. To verify the correctness of the new implementation, comparisons are performed with the analytical solution for the cell importance. Using the new generator, differences between Monte Carlo generated and analytical importances are less than 0.1%. Also, assumptions implicit in the original MCNP generator are shown to be poor in problems with high scattering media. The new generator is fully compatible with MCNP`s AVATAR{trademark} automatic variance reduction method. The new generator applications, together with AVATAR, gives MCNP an enhanced suite of variance reduction methods. The flexibility and efficacy of this suite is demonstrated in a neutron porosity tool well-logging problem.
Date: December 31, 1997
Creator: Evans, T.M. & Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

MCNP{trademark} Software Quality Assurance plan

Description: MCNP is a computer code that models the interaction of radiation with matter. MCNP is developed and maintained by the Transport Methods Group (XTM) of the Los Alamos National Laboratory (LANL). This plan describes the Software Quality Assurance (SQA) program applied to the code. The SQA program is consistent with the requirements of IEEE-730.1 and the guiding principles of ISO 900.
Date: April 1, 1996
Creator: Abhold, H.M. & Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

MCNP4B{sup {trademark}} verification and validation

Description: Several new features and bug fixes have been incorporated into the new release of MCNP. As required by the MCNP Software Quality Assurance Plan, these changes to the code and the test set are documented here for user reference. This document summarizes the new MCNP4B features and corrections, separated into major and minor groupings. Also included are a code cleanup section and a section delineating problems identified in LA-12839 which have not been corrected. Finally, we document the MCNP4B test set modifications and explain how test set coverage has been improved.
Date: August 1, 1996
Creator: Hendricks, J.S. & Court, J.D.
Partner: UNT Libraries Government Documents Department

AN ASSESSMENT OF MCNP WEIGHT WINDOWS

Description: The weight window variance reduction method in the general-purpose Monte Carlo N-Particle radiation transport code MCNPTM has recently been rewritten. In particular, it is now possible to generate weight window importance functions on a superimposed mesh, eliminating the need to subdivide geometries for variance reduction purposes. Our assessment addresses the following questions: (1) Does the new MCNP4C treatment utilize weight windows as well as the former MCNP4B treatment? (2) Does the new MCNP4C weight window generator generate importance functions as well as MCNP4B? (3) How do superimposed mesh weight windows compare to cell-based weight windows? (4) What are the shortcomings of the new MCNP4C weight window generator? Our assessment was carried out with five neutron and photon shielding problems chosen for their demanding variance reduction requirements. The problems were an oil well logging problem, the Oak Ridge fusion shielding benchmark problem, a photon skyshine problem, an air-over-ground problem, and a sample problem for variance reduction.
Date: January 1, 2000
Creator: HENDRICKS, J. S. & CULBERTSON, C. N.
Partner: UNT Libraries Government Documents Department

Recent MCNP developments

Description: MCNP is a widely used and actively developed Monte Carlo radiation transport code. Many important features have recently been added and more are under development. Benchmark studies not only indicate that MCNP is accurate but also that modern computer codes can give answers basically as accurate as the physics data that goes in them. Even deep penetration problems can be correct to within a factor of two after 10 to 25 mean free paths of penetration. And finally, Monte Carlo calculations, once thought to be too expensive to run routinely, can now be run effectively on desktop computers which compete with the supercomputers of yesteryear. 21 refs., 3 tabs.
Date: January 1, 1991
Creator: Hendricks, J.S. & Briesmeister, J.F.
Partner: UNT Libraries Government Documents Department

Monte Carlo techniques for analyzing deep penetration problems

Description: A review of current methods and difficulties in Monte Carlo deep-penetration calculations is presented. Statistical uncertainty is discussed, and recent adjoint optimization of splitting, Russian roulette, and exponential transformation biasing is reviewed. Other aspects of the random walk and estimation processes are covered, including the relatively new DXANG angular biasing technique. Specific items summarized are albedo scattering, Monte Carlo coupling techniques with discrete ordinates and other methods, adjoint solutions, and multi-group Monte Carlo. The topic of code-generated biasing parameters is presented, including the creation of adjoint importance functions from forward calculations. Finally, current and future work in the area of computer learning and artificial intelligence is discussed in connection with Monte Carlo applications. 29 refs.
Date: January 1, 1985
Creator: Cramer, S.N.; Gonnord, J. & Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

Benchmark analysis of MCNP{trademark} ENDF/B-VI iron

Description: The MCNP ENDF/B-VI iron cross-section data was subjected to four benchmark studies as part of the Hiroshima/Nagasaki dose re-evaluation for the National Academy of Science and the Defense Nuclear Agency. The four benchmark studies were: (1) the iron sphere benchmarks from the Lawrence Livermore Pulsed Spheres; (2) the Oak Ridge National Laboratory Fusion Reactor Shielding Benchmark; (3) a 76-cm diameter iron sphere benchmark done at the University of Illinois; (4) the Oak Ridge National Laboratory Benchmark for Neutron Transport through Iron. MCNP4A was used to model each benchmark and computational results from the ENDF/B-VI iron evaluations were compared to ENDF/B-IV, ENDF/B-V, the MCNP Recommended Data Set (which includes Los Alamos National Laboratory Group T-2 evaluations), and experimental data. The results show that the ENDF/B-VI iron evaluations are as good as, or better than, previous data sets.
Date: December 1, 1994
Creator: Court, J. D. & Hendricks, J. S.
Partner: UNT Libraries Government Documents Department

Performance of MCNP4A on seven computing platforms

Description: The performance of seven computer platforms has been evaluated with the MCNP4A Monte Carlo radiation transport code. For the first time we report timing results using MCNP4A and its new test set and libraries. Comparisons are made on platforms not available to us in previous MCNP timing studies. By using MCNP4A and its 325-problem test set, a widely-used and readily-available physics production code is used; the timing comparison is not limited to a single ``typical`` problem, demonstrating the problem dependence of timing results; the results are reproducible at the more than 100 installations around the world using MCNP; comparison of performance of other computer platforms to the ones tested in this study is possible because we present raw data rather than normalized results; and a measure of the increase in performance of computer hardware and software over the past two years is possible. The computer platforms reported are the Cray-YMP 8/64, IBM RS/6000-560, Sun Sparc10, Sun Sparc2, HP/9000-735, 4 processor 100 MHz Silicon Graphics ONYX, and Gateway 2000 model 4DX2-66V PC. In 1991 a timing study of MCNP4, the predecessor to MCNP4A, was conducted using ENDF/B-V cross-section libraries, which are export protected. The new study is based upon the new MCNP 25-problem test set which utilizes internationally available data. MCNP4A, its test problems and the test data library are available from the Radiation Shielding and Information Center in Oak Ridge, Tennessee, or from the NEA Data Bank in Saclay, France. Anyone with the same workstation and compiler can get the same test problem sets, the same library files, and the same MCNP4A code from RSIC or NEA and replicate our results. And, because we report raw data, comparison of the performance of other compute platforms and compilers can be made.
Date: December 31, 1994
Creator: Hendricks, J. S. & Brockhoff, R. C.
Partner: UNT Libraries Government Documents Department

A new MCNP{trademark} test set

Description: The MCNP test set is used to test the MCNP code after installation on various computer platforms. For MCNP4 and MCNP4A this test set included 25 test problems designed to test as many features of the MCNP code as possible. A new and better test set has been devised to increase coverage of the code from 85% to 97% with 28 problems. The new test set is as fast as and shorter than the MCNP4A test set. The authors describe the methodology for devising the new test set, the features that were not covered in the MCNP4A test set, and the changes in the MCNP4A test set that have been made for MCNP4B and its developmental versions. Finally, new bugs uncovered by the new test set and a compilation of all known MCNP4A bugs are presented.
Date: September 1, 1994
Creator: Brockhoff, R. C. & Hendricks, J. S.
Partner: UNT Libraries Government Documents Department

MCNP{trademark} ENDF/B-VI iron benchmark calculations

Description: Four iron shielding benchmarks have been calculated for, we believe the first time, with MCNP4A and its new ENDF/B-VI library. These calculations are part of the Hiroshima/Nagasaki dose re-evaluation for the National Academy of Sciences and the Defense Nuclear Agency. We believe these calculations are significant because they validate MCNP and the new ENDF/B-VI libraries. These calculations are compared to ENDF/B-V, experiment, and in some cases the recommended MCNP data library (a T-2 evaluation) and ENDF/IV.
Date: August 1, 1994
Creator: Court, J. D. & Hendricks, J. S.
Partner: UNT Libraries Government Documents Department

Comparison of neutron lifetimes as predicted by MCNP and DANTSYS

Description: The prompt removal lifetime algorithm used in the latest version of MCNP was modified to conform with the neutron-balance definitions described by Spriggs et al. In accordance with the neutron-balance theory, the non-adjoint-weighted removal lifetime is given by where {Phi} is the angular neutron flux, v is the neutron velocity, {Sigma}{sub a} is the macroscopic absorption cross section, E is neutron energy, {Omega} is angle, and r is a spatial vector. The numerator in this expression represents the total neutron population in the system, N, and the denominator represents the total loss rate due to leakage and absorption.
Date: January 22, 1997
Creator: Hendricks, J.S.; Parsons, D.K. & Spriggs, G.D.
Partner: UNT Libraries Government Documents Department

Definition of neutron lifespan and neutron lifetime in MCNP4B

Description: MCNP4B was released in early 1997. In this new version, several major changes were made to the underlying theory used to estimate the non-adjoint-weighted removal, fission, capture, and escape prompt-neutron lifetimes. These four lifetimes are now being calculated in accordance to the neutron-balance theory described by Spriggs et al. in which the non-adjoint-weighted lifetime for a particular type of reaction (i.e., fission, capture, escape, removal, etc.) is defined as the total neutron population in the system divided by that reaction rate.
Date: January 15, 1997
Creator: Busch, R. D.; Spriggs, G. D. & Hendricks, J. S.
Partner: UNT Libraries Government Documents Department