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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

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


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
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 next-event estimates from thermal collisions

Description: A new approximate method has been developed by Richard E. Prael to allow S({alpha},{beta}) thermal collision contributions to next-event estimators in Monte Carlo calculations. The new technique is generally applicable to next-event estimator contributions from any discrete probability distribution. The method has been incorporated into Version 4 of the production Monte Carlo neutron and photon radiation transport code MCNP. 9 refs.
Date: January 1, 1990
Creator: Hendricks, J.S. & Prael, R.E.
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

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

MCNP Perturbation Capability for Monte Carlo Criticality Calculations

Description: The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k{sub eff} in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward.
Date: September 20, 1999
Creator: Hendricks, J.S.; Carter, L.L. & McKinney, G.W.
Partner: UNT Libraries Government Documents Department

Installation of MCNP on 64-bit parallel computers

Description: The Monte Carlo radiation transport code MCNP has been successfully ported to two 64-bit workstations, the SGI and DEC Alpha. We found the biggest problem for installation on these machines to be Fortran and C mismatches in argument passing. Correction of these mismatches enabled, for the first time, dynamic memory allocation on 64-bit workstations. Although the 64-bit hardware is faster because 8-bytes are processed at a time rather than 4-bytes, we found no speed advantage in true 64-bit coding versus implicit double precision when porting an existing code to the 64-bit workstation architecture. We did find that PVM multiasking is very successful and represents a significant performance enhancement for scientific workstations.
Date: September 1, 1995
Creator: Meginnis, A.B.; Hendricks, J.S. & McKinney, G.W.
Partner: UNT Libraries Government Documents Department

Verification of the MCNP (TM) Perturbation Correction Feature for Cross-Section Dependent Tallies

Description: The Monte Carlo N-Particle Transport Code MCNP version 4B perturbation capability has been extended to cross-section dependent tallies and to the track-length estimate of Iqff in criticality problems. We present the complete theory of the MCNP perturbation capability including the correction to MCNP4B which enables cross-section dependent perturbation tallies. We also present the MCNP interface as an upgrade to the MCNP4B manual. Finally, we present test results demonstrating the validity of the perturbation capability in MCNP, particularly cross-section dependent problems.
Date: October 1, 1998
Creator: Hess, A. K.; McKinney, G. W.; Hendricks, J. S. & Carter, L. L.
Partner: UNT Libraries Government Documents Department

MCNP: Photon benchmark problems

Description: The recent widespread, markedly increased use of radiation transport codes has produced greater user and institutional demand for assurance that such codes give correct results. Responding to these pressing requirements for code validation, the general purpose Monte Carlo transport code MCNP has been tested on six different photon problem families. MCNP was used to simulate these six sets numerically. Results for each were compared to the set's analytical or experimental data. MCNP successfully predicted the analytical or experimental results of all six families within the statistical uncertainty inherent in the Monte Carlo method. From this we conclude that MCNP can accurately model a broad spectrum of photon transport problems. 8 refs., 30 figs., 5 tabs.
Date: September 1, 1991
Creator: Whalen, D.J.; Hollowell, D.E. & Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

MCNP neutron benchmarks

Description: Over 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The new and significant aspects of this work are as follows: These calculations are the first attempt at a validation program for MCNP and the first official benchmarking of version 4 of the code. We believe the chosen set of benchmarks is a comprehensive set that may be useful for benchmarking other radiation transport codes and data libraries. These calculations provide insight into how well neutron transport calculations can be expected to model a wide variety of problems.
Date: October 8, 1991
Creator: Hendricks, J.S.; Whalen, D.J.; Cardon, D.A. & Uhle, J.L.
Partner: UNT Libraries Government Documents Department