MCNP-REN - A Monte Carlo Tool for Neutron Detector Design Without Using the Point Model PDF Download
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Author: Publisher: ISBN: Category : Languages : en Pages : 5
Book Description
The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.
Author: Publisher: ISBN: Category : Languages : en Pages : 5
Book Description
The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.
Author: John S. Hendricks Publisher: Springer Nature ISBN: 3031041291 Category : Science Languages : en Pages : 316
Book Description
This open access book is a pedagogical, examples-based guide to using the Monte Carlo N-Particle (MCNP®) code for nuclear safeguards and non-proliferation applications. The MCNP code, general-purpose software for particle transport simulations, is widely used in the field of nuclear safeguards and non-proliferation for numerous applications including detector design and calibration, and the study of scenarios such as measurement of fresh and spent fuel. This book fills a gap in the existing MCNP software literature by teaching MCNP software usage through detailed examples that were selected based on both student feedback and the real-world experience of the nuclear safeguards group at Los Alamos National Laboratory. MCNP input and output files are explained, and the technical details used in MCNP input file preparation are linked to the MCNP code manual. Benefiting from the authors’ decades of experience in MCNP simulation, this book is essential reading for students, academic researchers, and practitioners whose work in nuclear physics or nuclear engineering is related to non-proliferation or nuclear safeguards. Each chapter comes with downloadable input files for the user to easily reproduce the examples in the text.
Author: Donald R. Culp Publisher: ISBN: Category : Languages : en Pages : 78
Book Description
Neutron transport calculations were performed using the Monte Carol code, MCNP. The transport problem considered has a point source at high altitude (40 km). Since atmospheric density decreases with altitude, two-dimensional effects (cylindrical coordinates) can be important. Results were compared to those obtained with the SMAUG-II computer code, which performs mass-integral scaling of approximate fits to one-dimensional spherical discrete ordinates solutions. These comparisons show the importance of two-dimensional computations. The report discusses practical issues of applying MCNP to this problem, without code modifications and includes example input files for MCNP and for SMAUG-II. Computational modelling of neutron transport in the atmosphere is complicated by the variation of air density with altitude.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.
Author: Publisher: ISBN: Category : Languages : en Pages : 7
Book Description
The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. The general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo Neutron Photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data. A rich collections of variance reduction features can greatly increase the efficiency of a calculation. MCNP is written in FORTRAN 77 and has been run on variety of computer systems from scientific workstations to supercomputers. The next production version of MCNP will include features such as continuous-energy electron transport and a multitasking option. Areas of ongoing research of interest to the well logging community include angle biasing, adaptive Monte Carlo, improved discrete ordinates capabilities, and discrete ordinates/Monte Carlo hybrid development. Los Alamos has requested approval by the Department of Energy to create a Radiation Transport Computational Facility under their User Facility Program to increase external interactions with industry, universities, and other government organizations. 21 refs.
Author: Publisher: ISBN: Category : Languages : en Pages : 5
Book Description
Nuclear safeguards active measurements that rely on the time correlation between fast neutrons and gamma rays from the same fission are a promising technique. Previous studies have shown the feasibility of this method, in conjunction with the use of artificial neural networks, to estimate the mass and enrichment of fissile samples enclosed in special, sealed containers. This paper evaluates the use of the associated particle sealed tube neutron generator (APSTNG) as the interrogation source in correlation measurements. The results show that its use is of particular importance when floor reflections are present. The Nuclear Materials Identification System (NMIS) presently uses 252Cf ionization chambers as interrogation sources for the time-dependent coincidence measurements. Because triggers from this source are associated with neutrons emitted in any direction, adjacent materials such as the floor and nearby containers could affect the measurements and should be accounted for. Conversely, the APSTNG, together with an alpha particle detector, defines a cone of neutrons that can be aimed at the item under verification, thus removing the effects of nearby materials from the time-dependent coincidence distributions. Monte Carlo calculations were performed using MCNP-POLIMI, a modified version of the standard MCNP code. The code attempts to calculate more correctly quantities that depend on the second moment of the neutron and gamma distributions. The simulations quantified the sensitivity enhancements and removal of the effects of nearby materials by substituting the traditional 252Cf source with the APSTNG.
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Author: John S. Hendricks
Author: Jian-Chyun Wang
Author: Svyatoslav Mihailovich Tkachenko
Author: Los Alamos Monte Carlo Group
Author: Donald R. Culp
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