| Ray Next-Event Estimator Transport of Primary and Secondary Gamma Rays |
Mar 2011 |
127 pages |
| Authors:
Whitman T Dailey; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING AND MANAGEMENT
|
 | This thesis investigated the application of the ray next event estimation Monte Carlo method to the transport of primary and secondary gamma rays. The problem of interest was estimation of the free field flux at a distant point in a vacuum from a point source in the atmosphere. An existing Fortran code for neutron transport, Ray Next-Event Estimator v4.0, was adapted to perform photon transport computations including coherent scattering, incoherent ... |
|
| Thermal Transitions and Reaction Kinetics of Polyhederal Silsesquioxane Containing Phenylethynylphthalimides (Preprint) |
18 Mar 2010 |
42 pages |
| Authors:
Bradley Seurer; Vandana Vij; Timothy Haddad; Joseph M Mabry; Andre Lee; AIR FORCE RESEARCH LAB EDWARDS AFB CA PROPULSION DIRECTORATE
|
| Thermal transitions and reaction kinetics of polyhedral oligomeric silsesquioxane (POSS) with a phenylethynylphthalimide (PEPI) moiety were investigated. Specifically, this study was aimed to understand the influence of the POSS periphery, types of spacer group in between the PEPI and the SiO core, architecture of PEPI arrangement with respect to the SiO core, and number of PEPI groups per cage on the thermal transitions and the crosslinking reaction of phynylethynyl. PEPI-POSS ... |
|
| Improving Low Order, Linear, Positive Spatial Quadratures for the Partial Current Neutron Transport Method |
Mar 2010 |
65 pages |
| Authors:
John M Snyder; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH GRADUATE SCHOOL OF ENGINEERING AND MANAGEMENT
|
| AFIT researchers have developed a new approach to solving Discrete Ordinates equations, which approximate the linear Boltzmann Transport Equation (BTE). The usual approach is von Neumann iteration on the scattering source, which requires repeated sweeps through the spatial-angular grid. Acceptable convergence requires complicated and expensive acceleration schemes. The new approach, Partial-Current Transport (PCT) with Adaptive Distribution Iteration, eliminates scattering source iteration through matrix inversions and a reduced-size global linear algebra ... |
|
| Effects of a Variable-Phase Transverse Acoustic Field on a Coaxial Injector at Subcritical and Near-Critical Conditions (Preprint) |
01-May-2008 |
12 pages |
| Authors:
Bruce Chehroudi; Douglas Talley; Juan I Rodriguez; Ivett A Leyva; AIR FORCE RESEARCH LAB EDWARDS AFB CA PROPULSION DIRECTORATE
|
| An experimental study that focuses on the effects of a variable transverse acoustic field on an N2 shear coaxial jet is presented. The coaxial jet is exposed to different acoustic conditions by varying the phase between two acoustic sources. The main objective of this investigation is to analyze the effect of transverse acoustic forcing with variable phase on the magnitude of the inner-jet dark-core length. The coaxial jet is exposed ... |
|
| Time Dependent Discrete Ordinates Neutron Transport Using Distribution Iteration in XYZ Geometry |
SEP 2007 |
146 pages |
| Authors:
James R. Dishaw; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH GRADUATE SCHOOL OF ENGINEERING AND MANAGEMENT
|
| The DI algorithm is an alternative to source iteration that, in our testing, does not require an accelerator. I developed a formal verification plan and executed it to verify the results produced by my code that implemented DI with the above features. A new, matrix albedo, boundary condition treatment was developed and implemented so that infinite-medium benchmarks could be included in the verification test suite. The DI algorithm was modified ... |
|
| Manufacture and Testing of an Activation Foil Package for Use in AFIDS |
MAR 2005 |
94 pages |
| Authors:
Warren E. Kimmel; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING AND MANAGEMENT
|
| This study used simulation and experiment to design and test foil packets for use in the Activation Foil Integrated Detection System (AFIDS). The initial plan to activate foil packets outside with the pulse reactor at White Sands Missile Range (WSMR) was not possible due to WSMR not having safety approval to take the reactor outside. As an alternative, the concept of using liquid nitrous oxide inside a reactor to simulate ... |
|
| Investigation of a Passive, Temporal, Neutron Monitoring System that Functions within the Confines of Start I |
MAR 2003 |
92 pages |
| Authors:
Stephanie Vaughn; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING AND MANAGEMENT
|
| This study is an investigation of the theoretical and experimental possibilities of using activation foils to detect and monitor special nuclear material for treaty monitoring purposes. None of the experiments demonstrated sufficient sensitivity to detect the target flux of 0.5 neutrons/cu cm--sec. The target flux could be detectable, if the limit of detection had been reduced by a factor of 4 to 6. However, many issues identified could enhance the ... |
|
| Research Report Point Reactor Kinetic Analysis |
DEC 2002 |
43 pages |
| Authors:
Daryl E. Neher Ii; ABERDEEN TEST CENTER ABERDEEN PROVING GROUND MD
|
| A computer code was written using a point reactor kinetics model Program results are compared to previous theoretical and APRF empirical pulse data. The program is used to determine temperature transients for different scram failures. The pulse-less tail mode of operation is discussed. |
|
| Efficient and Accurate Computation of Non-Negative Anisotropic Group Scattering Cross Sections for Discrete Ordinates and Monte Carlo Radiation Transport |
JUL 2002 |
106 pages |
| Authors:
David W. Gerts; AIR FORCE INST OF TECH WRIGHT-PATTERSONAFB OH SCHOOL OF ENGINEERING
|
| A new method for approximating anisotropic, multi-group scatter cross sections for use in discretized and Monte Carlo multi-group neutron transport is presented. The new method eliminates unphysical artifacts such as negative group scatter cross sections and falsely positive cross sections. Additionally, when combined with the discrete elements angular quadrature method, the new cross sections eliminate the lack of angular support in the discrete ordinates quadrature method. The new method generates ... |
|
| Electronic Reliability and the Environmental Thermal Neutron Flux |
06 MAY 2002 |
95 pages |
| Authors:
John D. Dirk; NAVAL ACADEMY ANNAPOLIS MD
|
| Boron, which is used in the manufacturing process of manufacturing, is highly sensitive to thermal neutrons. When ambient thermal neutrons originating from cosmic rays interact with the nucleus of boron, ionizing radiation is produced that can change the logic state of a cell on a microchip. This phenomenon is known as a Single Event Upset or Soft Error and is an important problem facing computer manufacturers. The goal of this ... |
|
| Computational Modeling of a Time-Independent, Heterogeneous Reactor Core Using Simplified Discrete Ordinates Neutron Transport Techniques |
SEP 2001 |
157 pages |
| Authors:
Kristofer S. Labowski; AIR FORCE INST OF TECH WRIGHT-PATTERSONAFB OH
|
| The Linear Characteristic (LC) method on rectangular boxoid meshes is a discrete ordinate neutron transport technique that uses both zeroth and first moments of the angular neutron flux to construct a relatively accurate representation of neutron particle distributions in a given medium. The significant number of calculations required by LC when compared to more conventional methods such as Diamond Difference vastly increases computational time by a factor of 60 (or ... |
|
| Dismantling Russia's Northern Fleet Nuclear Submarines: Environmental and Proliferation Risks |
JUN 2000 |
87 pages |
| Authors:
Benjamin A. Snell; NAVAL POSTGRADUATE SCHOOL MONTEREY CA
|
| This thesis examines the 1986 Chernobyl accident and its consequences as the basis for an analysis of the possible dimensions of the nuclear catastrophes that could occur during the dismantlement process of Russia's Northern Fleet nuclear submarines. It assesses the potential demographic, ecological, and economic consequences of a nuclear accident. Given the systemic problems at Russian nuclear facilities, the risks of a catastrophic event in the ... |
|
| An Innovative Computational Method for Predicting Neutron Transport |
MAR 2000 |
60 pages |
| Authors:
Jay I. Frankel; TENNESSEE UNIV KNOXVILLE COLL OF ENGINEERING
|
| This final report indicates the work performed under DNA-OO1-96-0103 during the span covering June 1996 through June 1998. Several important findings are reported strongly suggesting that the proposed numerical methodology can perform as initially conceived upon resolving the remaining technical issue involving the numerical integration of integral containing Hadamard type kernels. Preliminary calculations are presented in bath cartesian and spherical geometries indicating that the symbolically ... |
|
| Positive Anisotropic Group Scattering Cross Sections for Radiation Transport |
MAY 1999 |
207 pages |
| Authors:
J. M. DelGrande; AIR FORCE INST OF TECH WRIGHT-PATTERSONAFB OH SCHOOL OF ENGINEERING
|
| In solving the Boltzmann transport equation, most discrete ordinates codes calculate the source term by first approximating the scattering cross section using a Legendre polynomial expansion. Such expansions are insufficient when scattering is anisotropic and the Legendre expansion is truncated prematurely. This can lead to nonphysical negative cross sections, negative source terms and negative angular fluxes. While negative sources are problematic for standard discrete ordinates methods leading to poor convergence ... |
|
| Development of a Discrete Ordinates Code System for Unstructured Meshes of Tetrahedral Cells, with Serial and Parallel Implementations |
NOV 1998 |
198 pages |
| Authors:
Rodney L. Miller; AIR FORCE INST OF TECH WRIGHT-PATTERSONAFB OH SCHOOL OF ENGINEERING
|
| A numerically stable, accurate, and robust form of the exponential characteristic (EC) method, used to solve the time-independent linearized Boltzmann Transport Equation, is derived using direct affine coordinate transformations on unstructured meshes of tetrahedra. This quadrature, as well as the linear characteristic (LC) spatial quadrature, is implemented in our transport code, called TETRAN. This code solves multi-group neutral particle transport problems with anisotropic scattering and was parallelized using High Performance ... |
|
| Back to the Future: The Historical, Scientific, Naval, and Environmental Case for Fission Fusion |
02 APR 1998 |
28 pages |
| Authors:
Wallace M. Manheimer; NAVAL RESEARCH LAB WASHINGTON DC FUNDAMENTAL PLASMA PROCESSES
|
| It is proposed that a return to fission fusion, and especially the development of the thorium cycle could be a means to revitalize magnetic fusion research. This work analyzes recent history, attempts to find the reason magnetic fusion research is in the shape it is in, and argues that an embrace of the hybrid could improve its prospects. Then it analyzes recent Tokamak results, concluding that a research Tokamak reactor, ... |
|
| Reactions of Isotopically Labeled Nitric Oxide (N15O) in a Gas Phase Corona Reactor |
23 DEC 1997 |
7 pages |
| Authors:
B. R. Locke; R. J. Clark; G. Sathiamoorthy; W. C. Finney; FLORIDA STATE UNIV TALLAHASSEE DEPT OFCHEMISTRY
|
| The chemical reactions of nitric oxide decomposition in a pulsed streamer corona discharge were studied through the use of isotopically labeled nitric oxide, N(15)O. Detection of the relative abundance of N(14)-N(14), N(15)- N(14), N(15)-N(15) and N(15)O using mass spectrometry indicate that decreasing the initial concentration of NO leads to less direct dissociation of N(15)O by the corona discharge in favor of direct dissociation of N(14)-N(14). It was found that even ... |
|
| Modeling of Subcritical Penetration into Sediments Due to Interface Roughness |
JUN 1997 |
8 pages |
| Authors:
Eric I. Thorsos; Darrell R. Jackson; John E. Moe; Kevin L. Williams; UNIV OF WASHINGTON SEATTLE APPLIED PHYSICS LAB
|
| Recent experimental results reveal acoustic penetration into sandy sediments at grazing angles below the critical angle. We have been investigating a mechanism for subcritical penetration based on scattering at a rough water- sediment interface. Using perturbation theory, a numerically tractable three- dimensional model has been developed for simulating experiments. Data-model comparisons show that interface roughness is a viable hypothesis for the observed subcritical penetration. ... |
|
| Pentran: A Parallel 3-D S Transport Code With Complete Phase Space Decomposition, Adaptive Differencing, and Iterative Solution Methods |
09 JAN 97 |
|
| Authors:
Glenn E. Sjoden; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH
|
 | This thesis is based on the development and testing of the PENTRAN (Parallel Environment Neutral particle TRANsport) discrete ordinates code, designed for distributed memory and distributed computing parallel environments. The code can iteratively solve complex problems in nuclear design with complete, automatic decomposition of the angular, energy, and spatial variables. Written in ANSI FORTRAN-77 using the new standard Message Passing Interface library, PENTRAN is 28,500 lines and solves multigroup, anisotropic, ... |
|
| Startup Control of the TOPAZ-II Space Nuclear Reactor |
SEP 96 |
|
| Authors:
Carl D. Astrin; NAVAL POSTGRADUATE SCHOOL MONTEREY CA
|
| The Russian designed and manufactured TOPAZ-II Thermionic Nuclear Space Reactor has been supplied to the Ballistic Missile Defense Organization for study as part of the TOPAZ International Program. A Preliminary Nuclear Safety Assessment investigated the readiness to use the TOPAZ-II in support of a Nuclear Electric Propulsion Space Test Mission (NEPSTP). Among the anticipated system modifications required for launching the TOPAZ-II system within safety goals is for a U.S. designed ... |
|
| The TRIGA Reactor Facility at the Armed Forces Radiobiology Research Institute: A Simplified Technical Description. revision |
JAN 94 |
22 pages |
| Authors:
Mark L. Moore; ARMED FORCES RADIOBIOLOGY RESEARCH INST BETHESDA MD
|
| This publication provides a simplified technical description of the TRIGA research reactor at AFRRI. Topics covered include general principles of reactor operation and a description of the TRIGA reactor and its unique features. |
|
| Exponential Characteristic Spatial Quadrature for Discrete Ordinates Neutral Particle Transport in Two-Dimensional Cartesian Coordinates |
SEP 93 |
|
| Authors:
Bryan M. Minor; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING
|
| The exponential characteristic spatial quadrature for discrete ordinates neutral particle transport with rectangular cells is developed. Numerical problems arising in the derivation required the development of exponential moment functions. These functions are used to remove indeterminant forms which can cause catastrophic cancellations. The EC method is positive and nonlinear. It conserves particles and satisfies first moment balance. Comparisons of the EC method's performance to other methods in optically thin and ... |
|
| Linear Characteristic Spatial Quadrature for Discrete Ordinates Neutral Particle Transport on Arbitrary Triangles |
JUN 93 |
135 pages |
| Authors:
Dennis J. Miller; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH
|
| A new discrete ordinates spatial quadrature for arbitrary triangular cells is derived and compared to its rectangular cell linear characteristic counterpart. The triangular mesh is more flexible, allowing curved surfaces and off-axis angles to be approximated with many fewer spatial cells. The triangle method is consistently more accurate on example problems tested here. Arbitrary orientation and size of the triangles allow non-patterned meshes to be developed which appears to ameliorate ... |
|
| Reactivity Estimation and Validation for the Control of Reactor Neutronic Power |
MAY 93 |
264 pages |
| Authors:
Charles S. Lasota; MASSACHUSETTS INST OF TECH CAMBRIDGE DEPT OF ELECTRICAL ENGINEERING
|
| From July 1986 to July 1991, a joint MIT-SNL research team developed a controller capable of safely raising reactor power by approximately five orders of magnitude in a few seconds. This controller was experimentally demonstrated on the MIT Research Reactor (MITR-II) as well as on the 'Sandia National Laboratories' Annular Core Research Reactor (ACRR). This controller's intended application is for the control of spacecraft nuclear reactors. However, it also has ... |
|
| Visualization and Investigation of Natural Circulation in a Vessel with Tube Bundle Configuration Using DPIDV and LDV Comparison of DPIDV and LDV Measurements, |
23 JUL 1992 |
|
| Authors:
R. Kulenovic; S. Roesler; M. Groll; STUTTGART UNIV (GERMANY F R)
|
| This paper deals with the investigation of single-phase natural convection flows of air in a tube bundle configuration. Nuclear reactor safety research forms the background for the experiments. In this context the examination of circulation flow is important regarding high pressure loss-of-coolant-accidents. Safety considerations make it necessary to determine the influence of the developing natural convection flows on the heating-up of the reactor core and the reactor pressure vessel installations, ... |
|
| Maximum Temperature Calculation and Operational Characteristics of Fuel Follower Control Rods for the AFRRI TRIGA Reactor Facility |
MAY 91 |
23 pages |
| Authors:
M. Forsbacka; M. Moore; ARMED FORCES RADIOBIOLOGY RESEARCH INST BETHESDA MD
|
| Operational requirements of the Armed Forces Radiobiology Research Institute (AFRRI) TRIGA reactor necessitate the implementation of fuel follower control rods (FFCR's). This technical report discusses the operational and safety aspects of FFCR installation. Thermalhydraulic analysis shows that FFCR's can be operated safely in the AFRRI TRIGA reactor. The maximum calculated fuel temperature at the maximum steady-state power level is well within the limit of the maximum safe operating temperature set ... |
|
| A Time Dependent Transport Equation Solver |
MAY 91 |
|
| Authors:
Lennard W. Lee Jr; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH
|
| A new time dependent neutron and photon transport code was developed. The code, FMP2DT (Finite element, Multigroup, P sub n, 2-Dimensional, Time dependent), was discretized in space using finite elements, discretized in energy using a multigroup approximation, and discretized in time using Euler' backward differencing scheme. Its angular flux dependency was discretized using spherical harmonics. A P sub 1 angular flux approximation allows some modeling of both anisotropic flux behavior ... |
|
| Thermal-Hydraulic Transient Analysis of a Packed Particle Bed Reactor Fuel Element |
JUN 90 |
|
| Authors:
William E. Casey; MASSACHUSETTS INST OF TECH CAMBRIDGE
|
| A model which describes the thermal-hydraulic behavior of a packed particle bed reactor fuel element is developed and compared to a reference standard. The model represents a step toward a thermal-hydraulic module for a real-time, autonomous reactor powder controller. The general configuration of the fuel element is a bed of small (diameter about 500 microns) fuel particles are packed between concentrically mounted retention cylinders referred to as frits. The momentum ... |
|
| Multi-Modular Nuclear Reactor Plant Simulation and Control |
MAY 89 |
|
| Authors:
Mather K. Waltrip; MASSACHUSETTS INST OF TECH CAMBRIDGE DEPT OF NUCLEAR ENGINEERING
|
| A new generation of nuclear power plants now being considered will likely incorporate a multi-modular design strategy, in which separate nuclear steam supply modules provide steam to an aggregate turbine-generator. Smaller reactor cores in each module allow for the implementation of advanced safety features with relative ease and economy. Use of one relatively large turbine- generator set should help the multi-modular nuclear power plant to capture economies of scale by ... |
|
| The Feasibility of Controlled Rate of Release of Energy by Nuclear Alpha-Emitters |
JUN 88 |
|
| Authors:
David A. Sparrow; INSTITUTE FOR DEFENSE ANALYSES ALEXANDRIA VA
|
| Energy is approximately one million times more densely stored in atomic nuclei than in chemical bonds. This has been exploited in nuclear reactors which can supply varying amounts of electric power, but which have a minimum size, and in radioisotope thermoelectric generators (RTGs) which can be small, but have very limited power range despite their large energy density. This paper looks at some possibilities for externally varying alpha decay rates, ... |
|
| An Analysis of Heat Transfer after Loss of Primary Coolant in the SP-100 Reactor System |
MAR 88 |
75 pages |
| Authors:
Donald W. Robbins; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING
|
| This study determines design guidelines for the SP-100 space reactor core cooling system after a loss of coolant accident. The Thermal Systems Analysis Code (TSAP) calculated the temperatures within the fuel assemblies as a result of the fuel decay heat. TSAP is a lumped-parameter network analysis code capable of performing radiative and conductive heat transfer analysis. The reactor core was assumed to void of coolant instantaneously following a LOCA. The ... |
|
| Customizing SNPSAM: Introducing a Secondary Coolant Loop |
MAR 88 |
|
| Authors:
Vaughn H. Standley; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING
|
| The purpose of this study was to add a secondary coolant loop model to the Space Nuclear Power Systems Analysis Model (SNPSAM). The heat rejection systems, including the TEM pump, energy conversion assembly, and radiator, are emphasized while the reactor model is de-emphasized. Modifications were made to the TEM pump and heat exchanger models. A subroutine that simulated the secondary was written and SNPSAM was modified to incorporate this subroutine. ... |
|
| The Transient Limits of the Fast Flux Test Facility with Respect to the Passive Safety Program for Liquid Metal Fast Breeder Reactors |
87 |
|
| Authors:
Carl R. Sovinec; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH
|
| Reviewing the history behind the transient limits, including how they are presently calculated, teaches what affects the Fast Flux Test Facility (FFTF) capabilities. The purpose of this thesis is, therefore, to give the passive safety planners a compact source of this information. It covers the original methodology for calculating the transient limits, the NRC's criticisms of it, and the subsequent revisions at a level through enough for a basic understanding. ... |
|
| A Comparison Study of the Eigenvalue Method for the Solution of the Transient Heat Conduction Equation |
JAN 86 |
|
| Authors:
David B. Gee; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH
|
| This is a comparison study of the abilities of the eigenvalue method as a numerical method in solving the transient heat conduction equation. The eigenvalue method was compared to five other numerical methods; Runge-Kutta, Gears, extrapolation, fully implicit, and Crank-Nicolson. These methods were used to solved three physical problems. The first is a two dimensional slap which takes advantage of the symmetry of the problem. The second is a the ... |
|
| Conceptual Design and Neutronics Analyses of a Fusion Reactor Blanket Simulation Facility |
86 |
|
| Authors:
Denis E. Beller; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH
|
| A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed to calculation methods because design deficiencies which could affect ... |
|
| Discrete Elements Method of Neutral Particle Transport |
OCT 1983 |
135 pages |
| Authors:
K. A. Mathews; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING
|
| A new 'discrete elements' (LN) transport method is derived and compared to the discrete ordinates SN method, theoretically and by numerical experimentation. The discrete elements method is more accurate than discrete ordinates and strongly ameliorates ray effects for the practical problems studied. The discrete elements method is shown to be more cost effective in terms of execution time with comparable storage to attain the same accuracy, for a one-dimensional test ... |
|
| Single Event Upset Phenomena from High Energy Neutrons |
01 DEC 1981 |
|
| Authors:
R. L. Pease; J. P. Raymond; MISSION RESEARCH CORP ALBUQUERQUE NM
|
| The susceptibility of circumvented and uncircumvented endoatmospheric and exoatmospheric systems to neutron induced SEU(Single Event Upsets) was considered in very general terms of nuclear weapon threat and system configuration. For both exiatmospheric and endoatmospheric uncircumvented systems it appears that transient upset effects will be dominated by effects of the prompt ionization pulse. In the worst case neutron-induced SEUs are comparable to the upset levels of the prompt ionization pulse. For ... |
|
| The Annual Report of the Institute of Atomic Energy (Selected Pages) |
12 AUG 1981 |
|
| Authors:
FOREIGN TECHNOLOGY DIV WRIGHT-PATTERSON AFB OH
|
|
| An Interactive Code for a Pressurized Water Reactor Incorporating Temperature and Xenon Feedback. |
JUN 1980 |
|
| Authors:
Gregory Garver Heath; NAVAL POSTGRADUATE SCHOOL MONTEREY CA
|
| An interactive computer model of a highly enriched pressurized water reactor was developed, using the applicable plant parameters from the Shippingport Atomic Power Station. The point reactor kinetics equations for one delayed neutron precursor group were linearized using small perturbation theory. The model included both moderator and Xenon-135 reactivity feedback effects, as well as an automatic reactor protection and average reactor coolant temperature control system. The thermal response of the ... |
|
| Monte Carlo Methods for Neutron Flux Calculations in a Pressurized Light Water Power Reactor Using Morse-CG. |
27 NOV 1979 |
|
| Authors:
R. A. Lindgren ; M. Rosen ; A. I. Namenson; NAVAL RESEARCH LAB WASHINGTON DC
|
| Various methods of optimizing Monte Carlo calculations of fast flux and spectra in a pressurized light water reactor are investigated using the code MORSE. Results in slab and cylindrical geometry are compared with those of the code ANISN. (Author) |
|
| Integral Equation Space-Energy Flux Synthesis for Spherical Systems. |
21 NOV 1979 |
|
| Authors:
Eugene W. Skluzacek; AIR FORCE TECHNICAL APPLICATIONS CENTER PATRICK AFB FL
|
| The calculations of neutron flux distribution and growth rate for small, spherically symmetric systems usually requires extensive computing time on the largest machines. To minimize computing time, a compromise between the simplicity of diffusion theory and the accuracy of transport theory is needed. The Serber-Wilson method, Feynman's method, and early flux synthesis methods are used as the foundation for integral equation synthesis (IES) which is an approximate, numerical technique for ... |
|
| Application of Phase-Space Finite Elements to the Neutron Transport Equation in Cylindrical Geometry. |
DEC 1978 |
|
| Authors:
Ronald C. Wheaton; AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OHIO SCHOOL OF ENGINEERING
|
| Phase-Space finite elements are applied to the static monoenergetic neutron transport equation in two-dimensional cylindrical geometry by computer subroutines which collectively assemble the global phase-space matrix for solution. The technique uses a variational formulation based on the second-order self-adjoint form of the transport equation within which the dependent variable approximated by the finite elements is the even-parity component of the angular flux. (Author) |
|
| A Sensitivity Study of the Critical Flow Models Used in RELAP4/MOD 5 Blowdown Analysis of a General Electric BWR. |
NOV 1978 |
|
| Authors:
Mark Vearil Carle; NAVAL POSTGRADUATE SCHOOL MONTEREY CA
|
| This study uses the RELAP4/MOD5 Computer Code to analyze the effects of critical flow modeling on the thermal-hydraulic transient response of a General Electric boiling water reactor to a major primary coolant line rupture. Included is a presentation of the equations, assumptions, and limitations of the critical flow models available for use in RFLAP4. Additionally, an evaluation of a temporary solution to a RELAP4 coding error associated with stagnation properties ... |
|
| Combined Effect of Aging and Neutron Irradiation on Semiconductor Avalanche Voltage. |
APR 1978 |
|
| Authors:
Victor W. Ruwe; ARMY MISSILE RESEARCH AND DEVELOPMENT COMMAND REDSTONE ARSENAL AL ENGINEERING LAB
|
| This report presents the results of an investigation into the combined effects of neutron irradiation and aging on the breakdown voltage in transistors. The combined effect was found to be a simple additive effect. It was determined from other results of the investigation that transistor parameters change as a function of the number of times the breakdown voltage is measured and that the gain of the transistor is degraded more ... |
|
| Operator-Valued Chandrasekhar H-Functions. |
DEC 1977 |
|
| Authors:
C. T. Kelley; WISCONSIN UNIV MADISON MATHEMATICS RESEARCH CENTER
|
| Operator valued analogs of the Chandrasekhar H-function that occur in the study of neutron transport in a slab with continuous energy dependence and anisotropic scattering satisfy a system of nonlinear integral equations. An appropriate Banach space setting is found for the study of this system. The system may be solved by iteration. The domain of analyticity of H sub r and H sub l is extended by means of bifurcation ... |
|
| Preliminary Study of High Neutron Flux Fusion Heating. |
03 NOV 1977 |
|
| Authors:
Richard L. Liboff; CORNELL UNIV ITHACA N Y SCHOOL OF ELECTRICAL ENGINEERING
|
| A heating scheme for nuclear fusion is proposed based on the availability of a high flux, low energy neutron source. The heat is derived in the reaction Li(6)(n,T)He(4) resulting from the incidence of a low energy neutron beam on a sample of Li(6)D. The energy release per reaction, Q = 4.6 MeV, is converted through a electron Coulomb collisions thereby quickly dissociating the solid sample to the plasma state. For ... |
|
| Fast Neutron Induced Activity of Elements Relevant to Fusion Reactor Structural Design. |
APR 1977 |
|
| Authors:
E. A. Kamykowski; GRUMMAN AEROSPACE CORP BETHPAGE N Y RESEARCH DEPT
|
| The designer of nuclear fusion devices must consider the problem of induced activation of system components resulting from irradiation by the generated neutron flux. Requirements for maintenance and repair, as well as for storage of exposed equipment, can be directly affected by the levels of activity induced and by the radio active decay characteristics of the constituent elements. Since the activation cross sections and decay properties of these elements have ... |
|
| SAMCEP: A Correlated Monte Carlo Neutron and Gamma Radiation Transport Code. |
FEB 1977 |
|
| Authors:
Henry Lichtenstein; Herbert Steinberg; Joan Brooks; MATHEMATICAL APPLICATIONS GROUP INC ELMSFORD N Y
|
| SAMCEP is a system of U.S.A. - Standard FORTRAN programs for the solution by Monte Carlo of several correlated neutron and secondary gamma transport problems in complex three-dimensional geometries. Up to 10 correlated problems, involving perturbations of composition, neutron cross sections, angular or energy distributions, and/or source spectra, as well as perturbations of gamma production data, can be run simultaneously. Individual problem fluxes (region-, energy-, and time-dependent), as well as ... |
|
| A Comparison of Integration Methods for the Solution of Nonlinear Reactor Dynamics Problems through the Use of Finite Elements. |
DEC 1976 |
|
| Authors:
Ralph Carroll Sheldrick; NAVAL POSTGRADUATE SCHOOL MONTEREY CALIF
|
| A comparison of numerical methods utilized by the finite element technique for solving a nonlinear nuclear reactor dynamics problem was conducted. Using the Crank-Nicolson, DVOGER (Gear) and Implicit Gear methods, the results showed the Implicit to be the superior method investigated. This is based on the fact that all three methods yielded the same steady state solutions; but, the Implicit Gear method used significantly less CPU time and comparable storage ... |
|
| Finite Element Solution of a Three-Dimensional Nonlinear Reactor Dynamics Problem with Feedback. |
DEC 1976 |
|
| Authors:
Eulogio Conception Bermudes; NAVAL POSTGRADUATE SCHOOL MONTEREY CALIF
|
| This work examines the three-dimensional dynamic response of a nonlinear fast reactor with temperature-dependent feedback and delayed neutrons when subjected to uniform and local disturbances. The finite element method was employed to reduce the partial differential reactor equation to a system of ordinary differential equations which can be numerically integrated. A program for the numerical solution of large sparse systems of stiff differential equations developed by Franke and based on ... |
|
Fission Reactor Physics
