37 Matching Results

Search Results

Advanced search parameters have been applied.

Monte Carlo calculations of neutron and gamma-ray energy spectra for fusion reactor shield design: comparison with experiment

Description: Neutron and gamma-ray energy spectra resulting from the interactions of approx. 14 MeV neutrons in laminated slabs of stainless steel type-304 and borated polyethylene have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data as a function of slab thickness and material composition and as a function of detector location behind the slabs. Comparisons of the differential energy spectra are made for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The measured neutron spectra and those calculated using Monte Carlo methods agree witin 5% to 50% depending on the slab thickness and composition and neutron energy. The agreement between the measured and calculated gamma-ray energy spectra are also within this range. The MCNP data are also in favorable agreement with attenuated data calculated previously by discrete ordinates transport methods and the Monte Carlo code SAM-CE.
Date: January 1, 1983
Creator: Santoro, R.T. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department

Nuclear design calculations for the TFTR neutral beam injectors

Description: The results of two-dimensional neutronics calculations aimed at identifying the nuclear responses that affect the performance of vital components are summarized. The neutral beam injector assembly was modelled using r-z geometry with cylindrical symmetry about z-axis. The dimensions and compositions of the components were taken from engineering drawings and sketches. The neutron source strength was assumed to be composed entirely of 14-MeV neutrons isotropically emitted from an area having a radius of 70 cm. The total neutron yield from the reactor was 7 x 10/sup 18/ neutrons/second. The radiation transport calculations were performed using the two-dimensional discrete ordinates code DOT with an S/sub 8/ angular quadrature, and a P/sub 3/ Legendre expansion of the cross section. (MHR)
Date: January 1, 1977
Creator: Santoro, R.T.; Alsmiller, R.G. Jr. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department

Plasma engineering analysis of an EBT operating window

Description: The operating space for EBT reactors is calculated using a newly developed systems code that incorporates recent advances in EBT physics. The calculation includes a self-consistent treatment of coupled ring-core stability and power balance requirements. The essential elements of the systems code are reviewed including magnetics, stability, ring power, power balance, confinement time, and cost calculations. Finally, a typical reactor systems analysis is summarized for a family of EBT reactors that fall within the allowed operating space.
Date: January 1, 1983
Creator: Santoro, R.T.; Uckan, N.A. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department

Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

Description: For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig.
Date: January 1, 1986
Creator: Alsmiller, R.G. Jr.; Barnes, J.M. & Drischler, J.D.
Partner: UNT Libraries Government Documents Department

Nuclear performance calculations for the ELMO Bumpy Torus Reactor (EBTR) reference design

Description: The nuclear performance of the ELMO Bumpy Torus Reactor reference design has been calculated using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV transport cross-section data and nuclear response functions. The calculated results include estimates of the spatial and integral heating rate with emphasis on the recovery of fusion neutron energy in the blanket assembly and minimization of the energy deposition rates in the cryogenic magnet coil assemblies. The tritium breeding ratio in the natural lithium-laden blanket was calculated to be 1.29 tritium nuclei per incident neutron. The radiation damage in the reactor structural material and in the magnet assembly is also given.
Date: December 1, 1977
Creator: Santoro, R. T. & Barnes, J. M.
Partner: UNT Libraries Government Documents Department

Nuclear accident dosimetry: calculations and comparison with experimental data

Description: Calculated results, carried out by the method of discrete ordinates, of the absorbed dose received by personnel at various locations relative to the source in a simulated reactor criticality accident are presented and compared with experimental data. The geometry of the room in which the ''accident'' took place is included approximately in the calculations. The calculated and experimental data are in good agreement in all cases.
Date: July 1, 1978
Creator: Santoro, R.T.; Alsmiller, R.G. Jr. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department

Nuclear accident dosimetry: calculations and comparison with experimental data

Description: Calculated results, carried out by the method of discrete ordinates, of the absorbed dose received by personnel at various locations relative to the source in a simulated reactor criticality accident are presented and compared with experimental data. The geometry of the room in which the ''accident'' took place is included approximately in the calculations. The calculated and experimental data are in good agreement in all cases.
Date: July 1, 1978
Creator: Santoro, R.T.; Alsmiller, R.G. Jr. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department

Integral experiments for fusion reactor design: analysis

Description: Integral experiments that measure the energy spectra of neutrons and gamma rays due to the transport of approx. 14 MeV D-T neutrons through laminated SS-304 and borated polyethylene shield assemblies have been performed. Measured and calculated energy spectra and integrated flux distributions are compared for a typical shield assembly as a function of detector location.
Date: January 1, 1979
Creator: Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M. & Oblow, E.M.
Partner: UNT Libraries Government Documents Department

Two-dimensional neutronics analysis of the Tokamak Fusion Test Reactor

Description: The nuclear performance of the TFTR has been studied using two-dimensional radiation transport methods. The calculations were carried out using the two-dimensional discrete ordinates code DOT using S/sub 8/ angular quadrature and a P/sub 3/ expansion of the transport cross-section data. The cross sections were taken from the DLC-37 library and collapsed to a 35-neutron--21-gamma-ray energy group subset to facilitate the calculations. The TFTR and the concrete pillbox-shaped shield that surrounds the reactor assembly were modelled using r-z geometry with toroidal symmetry about the z-axis. The components were described using 61 radical and 38 axial mesh intervals. The dimensions and spatial location of the D-T plasma, the dimensions and compositions of the reactor components and other relevant TFTR parameters are summarized. (MHR)
Date: January 1, 1977
Creator: Santoro, R.T.; Barnes, J.M.; Lillie, R.A. & Alsmiller, R.G. Jr.
Partner: UNT Libraries Government Documents Department

Overview of the NSNS target station

Description: The technologies that are being utilized to design and build a state-of-the-art neutron spallation source, the National Spallation Neutron Source (NSNS), are discussed. Emphasis is given to the technology issues that present the greatest scientific challenges. The present facility configuration, ongoing analysis and the planned hardware research and development program are also described.
Date: April 1, 1997
Creator: Gabriel, T.A.; Barnes, J.M. & Charlton, L. A.
Partner: UNT Libraries Government Documents Department

Pre-conceptual design and preliminary neutronic analysis of the proposed National Spallation Neutron Source (NSNS)

Description: The Department of Energy (DOE) has initiated a pre-conceptual design study for the National Spallation Neutron Source (NSNS) and given preliminary approval for the proposed facility to be built at Oak Ridge National Laboratory (ORNL). The pre-conceptual design of the NSNS initially consists of an accelerator system capable of delivering a 1 to 2 GeV proton beam with 1 MW of beam power in an approximate 0.5 {micro}s pulse at a 60 Hz frequency onto a single target station. The NSNS will be upgradable to a significantly higher power level with two target stations (a 60 Hz station and a 10 Hz station). There are many possible layouts and designs for the NSNS target stations. This paper gives a brief overview of the proposed NSNS with respect to the target station, as well as the general philosophy adopted for the neutronic design of the NSNS target stations. A reference design is presented, and some preliminary neutronic results for the NSNS are briefly discussed.
Date: March 1, 1997
Creator: Johnson, J.O.; Barnes, J.M. & Charlton, L.A.
Partner: UNT Libraries Government Documents Department

Development of the activation analysis calculational methodology for the Spallation Neutron Source (SNS)

Description: For the design of the proposed Spallation Neutron Source (SNS), activation analyses are required to determine the radioactive waste streams, on-line material processing requirements remote handling/maintenance requirements, potential site contamination and background radiation levels. For the conceptual design of the SNS, the activation analyses were carried out using the high-energy transport code HETC96 coupled with MCNP to generate the required nuclide production rates for the ORIHET95 isotope generation code. ORIHET95 utilizes a matrix-exponential method to study the buildup and decay of activities for any system for which the nuclide production rates are known. In this paper, details of the developed methodology adopted for the activation analyses in the conceptual design of the SNS are presented along with some typical results of the analyses.
Date: March 1, 1998
Creator: Odano, N.; Johnson, J.O.; Charton, L.A. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department

SNS moderator design

Description: The pulsed-neutron source SNS facility will start operation at 1 MW. A later upgrade to 5 MW is planned. The facility consists of a linear accelerator, an accumulator ring, and a target station. The protons from the accumulator ring will be injected into the target station at 1 GeV. The subsequent spallation process will then produce low-energy thermal neutrons that may be used for a wide variety of experiments. In this paper the authors discuss neutronic calculations which address various aspects of the moderate design. The computer codes HETC and MCNP were used for these calculations with the former code performing the high-energy transport. Neutrons which fell in energy to 20 MeV or less were then passed to MCNP for further transport.
Date: December 1, 1997
Creator: Charlton, L.A.; Barnes, J.M.; Gabriel, T.A. & Johnson, J.O.
Partner: UNT Libraries Government Documents Department

Analysis of the radiation fallout tests at ETBS, France (Fall 1996)

Description: A series of experiments were carried out at the Etablissement Technique de Bourges (ETBS), France to measure protection factors for the Russian T72M tank during exposure to gamma radiation emanating from the ground. The purpose of these measurements was to determine the reduction in the dose rate to the tank occupants when the vehicle traverses terrain that has been contaminated as the result of fallout from a nuclear weapon or when the ground has been contaminated by the distribution of radioactive material by terrorists. This report summarizes results of calculations that replicate the measurements. Comparisons of measured and calculated protection factors are reported for a series of nested iron cylinders and the T72M tank. The cylinder measurements were performed to compare protection factors measured at Bourges with those obtained previously at the US Army Aberdeen Test Center.
Date: January 1, 2000
Creator: Barnes, J.M. & Santoro, R.T.
Partner: UNT Libraries Government Documents Department

Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

Description: Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method.
Date: August 1, 1980
Creator: Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M. & Chapman, G.T.
Partner: UNT Libraries Government Documents Department

Comparison of calculated and experimental neutron attenuation and streaming data for fusion reactor design

Description: Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n)/sup 4/He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct (L/D = 3) imbedded in a concrete shield have been performed at the Oak Ridge National Laboratory. The facility, NE-213 liquid scintillator detector system, and experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed.
Date: January 1, 1980
Creator: Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M. & Chapman, G.T.
Partner: UNT Libraries Government Documents Department

Sensitivity of the first-wall radiation damage to fusion reactor blanket composition

Description: The atomic displacement and hydrogen and helium gas production rates in a 1-cm-thick type-316 stainless steel first wall have been calculated as a function of blanket composition in a typical one-dimensional fusion reactor model. For a 50-cm-thick blanket, variations in the rates of atomic displacement and hydrogen and helium gas production of factors of 2.7, 1.3, and 1.2, respectively, were obtained. The dependence of the radiation damage responses on the thickness of the first wall and blanket are also given.
Date: December 1, 1977
Creator: Barnes, J. M.; Santoro, R. T. & Gabriel, T. A.
Partner: UNT Libraries Government Documents Department

Monte Carlo and discrete ordinates calculations of 14-MeV neutrons streaming through a stainless-steel duct (L/D = 4. 6): comparison with experiment

Description: Integral experiments are being carried out at the Oak Ridge National Laboratory to measure neutron and gamma-ray energy spectra from approx. 14 MeV neutrons streaming through cylindrical ducts imbedded in concrete. These data are being used to validate radiation transport code and nuclear data libraries that are being used to calculate nuclear streaming through the penetrations that are found in fusion reactor blanket and shield assemblies. In this paper, measured and calculated neutron and gamma-ray energy spectra from approx. 14 MeV neutrons streaming through a stainless-steel duct having a length-to-dia ratio of 4.6 are compared as a function of detector location relative to the mouth of the duct. The length of the duct is 1.45 m.
Date: January 1, 1985
Creator: Santoro, R.T.; Barnes, J.M.; Alsmiller, R.G. Jr. & Drischler, J.D.
Partner: UNT Libraries Government Documents Department

Neutron and gamma-ray streaming calculations for the ETF neutral-beam injectors

Description: The tritium plasma of the Engineering Test Facility (ETF) fusion reactor will be heated and ignited by the injection of neutral deuterium. Since the deuterons must be injected through straight ducts into the plasma, the neutron and secondary gamma radiation produced as a result of the D-T reactions will stream directly into the neutral beam injectors and lead to adverse effects in vital components. The radiation leaking through the injection ports will be comprised of approx. 14 MeV neutrons (from the D-T reactions) plus a low-energy neutron and secondary gamma ray distribution that results from the interactions of the energetic neutrons with the plasma liner and the primary shielding about the torus. In this paper two-dimensional radiation transport calculations carried out to estimate the effects on the injector components of radiation streaming through the injection duct will be described and the results of these calculations will be presented and discussed.
Date: January 1, 1981
Creator: Lillie, R.A.; Santoro, R.T.; Alsmiller, R.G. Jr. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department

The High-Energy Transport Code HETC88

Description: An upgraded version, HETC88, of the previously available High-Energy Transport Code HETC is briefly described. In the upgraded code, the particle production model from hadron-nucleus nonelastic collisions at energies greater than 5 GeV has been revised. At nucleon and ion energies below 5 GeV, HETC88 is not different from the code previously available. In particular, provision is still made to allow neutrons with energies less than or equal to20 MeV to be transported by none of the available codes designed for low-energy neutron transport. Calculated results for the longitudinal distribution of the flux of neutrons with energy greater than or equal to40 KeV in the Tevatron tunnel when 900 GeV protons interact with N/sub 2/ in a warm section are presented and compared with experimental data. Some disagreements between the calculated and measured neutron flux are found. For 20 TeV protons incident on a large cylindrical iron target, calculated ''star'' density results from HETC88, FLUKA87, CASIM, and MARS10 are also compared. 22 refs., 3 figs.
Date: January 1, 1989
Creator: Alsmiller, R.G. Jr.; Alsmiller, F.S.; Gabriel, T.A.; Hermann, O.W. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department

Integral experiments for fusion-reactor shield design. Summary of progress

Description: Neutron and gamma-ray energy spectra from the reactions of approx. 14-MeV neutrons in blanket and shield materials and from the streaming of these neutrons through a cylindrical duct (L/D approx. 2) have been measured and calculated. These data are being obtained in a series of integral experiments to verify the radiation transport methods and nuclear data that are being used in nuclear design calculations for fusion reactors. The experimental procedures and analytical methods used to obtain the calculated data are reviewed. Comparisons between measured and calculated data for the experiments that have been performed to date are summarized.
Date: January 1, 1983
Creator: Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M. & Chapman, G.T.
Partner: UNT Libraries Government Documents Department

Shielding calculations for the TFTR neutral beam injectors

Description: Two-dimensional discrete ordinates calculations have been performed to determine the location and thickness of concrete shielding around the Tokamak Fusion Test Reactor (TFTR) neutral beam injectors. Two sets of calculations were performed: one to determine the dose equivalent rate on the roof and walls of the test cell building when no injectors are present, and one to determine the contribution to the dose equivalent rate at these locations from radiation streaming through the injection duct. Shielding the side and rear of the neutral beam injector with 0.305 and 0.61 m of concrete, respectively, and lining the inside of the test cell wall with an additional layer of concrete having a thickness of 0.305 m and a height above the axis of deuteron injection of 3.10 m are sufficient to maintain the biological dose equivalent rate outside the test cell to approx. 1 mrem/DT pulse.
Date: July 1, 1979
Creator: Santoro, R.T.; Lillie, R.A.; Alsmiller, R.G. Jr. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department

Analysis of the ORNL radiation-streaming integral experiments

Description: The determination of radiation streaming through the various penetrations that will be present in the blanket and shield of a D-T burning fusion reactor is one of the more complicated problems facing nuclear design engineer. Penetrations of varying size and shape will be required for neutral particle injection, rf heating, vacuum pumping, and plasma diagnostics. The radiation streaming through these openings will degrade the reactor performance by the adverse effects of nuclear heating, induced activation, and radiation damage in vital components. The effects of radiation streaming must be estimated during the design of the reactor. Measurements have been made to determine the streaming of approx. 14-MeV neutrons through a simple iron duct imbedded in a concrete shield.
Date: January 1, 1981
Creator: Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T. & Tang, J.S.
Partner: UNT Libraries Government Documents Department

Two- and three-dimensional neutronics analyses of the TFTR neutral beam injectors

Description: The two-dimensional discrete ordinates code DOT using an S/sub 8/ angular quadrature and the three-dimensional Monte Carlo code MORSE were each used to calculate the spatial dependences of the nuclear heating rates and neutron and gamma-ray scalar flux distributions in the injector. In the three-dimensional analysis, these data were estimated at several locations in a detailed model of the injector using accurate neutron and gamma-ray energy, spatial, and angular distributions to describe the radiation sources incident on the injector. (MOW)
Date: January 1, 1978
Creator: Santoro, R.T.; Lillie, R.A.; Alsmiller, R.G. Jr. & Barnes, J.M.
Partner: UNT Libraries Government Documents Department