Neutron fluence rate and energy spectrum in SPRR-300 reactor thermal column

International Nuclear Information System (INIS)

Dou Haifeng; Dai Junlong

2006-01-01

In order to modify the simple one-dimension model, the neutron fluence rate distribution calculated with ANISN code ws checked with that calculated with MCNP code. To modify the error caused by ignoring the neutron landscape orientation leaking, the reflector that can't be modeled in a simple one-dimension model was dealt by extending landscape orientation scale. On this condition the neutron fluence rate distribution and the energy spectrum in the thermal column of SPRR-300 reactor were calculated with one-dimensional code ANISN, and the results of Cd ratio are well accorded with the experimental results. The deviation between them is less than 5% and it isn't above 10% in one or two special positions. It indicates that neutron fluence rate distribution and energy spectrum in the thermal column can be well calculated with one-dimensional code ANISN. (authors)

Seismically constrained two-dimensional crustal thermal structure of ...

Indian Academy of Sciences (India)

The temperature field within the crust is closely related to tectonic history as well as many other geological processes inside the earth. Therefore, knowledge of the crustal thermal structure of a region is of great importance for its tectonophysical studies. This work deals with the two-dimensional thermal modelling to ...

Thermal analysis of Ti drive-in target for D-D neutron generation

International Nuclear Information System (INIS)

Jung, N.S.; Kim, I.J.; Kim, S.J.; Choi, H.D.

2008-01-01

Full text: Thermal analysis was performed for a Ti drive-in target of a D-D neutron generator. Numerical calculation was the only feasible way to obtain the information of the target temperature, since it was very difficult to measure the target temperature during neutron generation due to high voltage being applied to the target. Computational fluid dynamics code CFX-5 was used in this study. In order to define the heat flux term for the thermal analysis, the current profile of the ion beam was measured. The one-dimensional, integrated current profile was measured by using a single slit and a Faraday cup. The measured current profile was transformed into the axially symmetric two-dimensional distribution function by using the Abel inversion, which had the two-dimensional Gaussian function shape. Temperature distribution in the target was calculated at the operating condition. The influence of operational parameters like the ion beam energy, current, coolant mass flow rate and coolant inlet temperature on the target temperature was investigated

Consideration of a ultracold neutron source in two-dimensional cylindrical geometry by taking simulated boundaries

Energy Technology Data Exchange (ETDEWEB)

Gheisari, R., E-mail: gheisari@pgu.ac.ir [Physics Department, Persian Gulf University, Bushehr 75169 (Iran, Islamic Republic of); Nuclear Energy Research Center, Persian Gulf University, Bushehr 75169 (Iran, Islamic Republic of); Firoozabadi, M. M.; Mohammadi, H. [Department of Physics, University of Birjand, Birjand 97175 (Iran, Islamic Republic of)

2014-01-15

A new idea to calculate ultracold neutron (UCN) production by using Monte Carlo simulation method to calculate the cold neutron (CN) flux and an analytical approach to calculate the UCN production from the simulated CN flux was given. A super-thermal source (UCN source) was modeled based on an arrangement of D{sub 2}O and solid D{sub 2} (sD{sub 2}). The D{sub 2}O was investigated as the neutron moderator, and sD{sub 2} as the converter. In order to determine the required parameters, a two-dimensional (2D) neutron balance equation written in Matlab was combined with the MCNPX simulation code. The 2D neutron-transport equation in cylindrical (ρ − z) geometry was considered for 330 neutron energy groups in the sD{sub 2}. The 2D balance equation for UCN and CN was solved using simulated CN flux as boundary value. The UCN source dimensions were calculated for the development of the next UCN source. In the optimal condition, the UCN flux and the UCN production rate (averaged over the sD{sub 2} volume) equal to 6.79 × 10{sup 6} cm{sup −2}s{sup −1} and 2.20 ×10{sup 5} cm{sup −3}s{sup −1}, respectively.

Solution of two-dimensional equations of neutron transport in 4P0-approximation of spherical harmonics method

International Nuclear Information System (INIS)

Polivanskij, V.P.

1989-01-01

The method to solve two-dimensional equations of neutron transport using 4P 0 -approximation is presented. Previously such approach was efficiently used for the solution of one-dimensional problems. New an attempt is made to apply the approach to solution of two-dimensional problems. Algorithm of the solution is given, as well as results of test neutron-physical calculations. A considerable as compared with diffusion approximation is shown. 11 refs

Criticality analysis of thermal reactors for two energy groups applying Monte Carlo and neutron Albedo method

International Nuclear Information System (INIS)

Terra, Andre Miguel Barge Pontes Torres

2005-01-01

The Albedo method applied to criticality calculations to nuclear reactors is characterized by following the neutron currents, allowing to make detailed analyses of the physics phenomena about interactions of the neutrons with the core-reflector set, by the determination of the probabilities of reflection, absorption, and transmission. Then, allowing to make detailed appreciations of the variation of the effective neutron multiplication factor, keff. In the present work, motivated for excellent results presented in dissertations applied to thermal reactors and shieldings, was described the methodology to Albedo method for the analysis criticality of thermal reactors by using two energy groups admitting variable core coefficients to each re-entrant current. By using the Monte Carlo KENO IV code was analyzed relation between the total fraction of neutrons absorbed in the core reactor and the fraction of neutrons that never have stayed into the reflector but were absorbed into the core. As parameters of comparison and analysis of the results obtained by the Albedo method were used one dimensional deterministic code ANISN (ANIsotropic SN transport code) and Diffusion method. The keff results determined by the Albedo method, to the type of analyzed reactor, showed excellent agreement. Thus were obtained relative errors of keff values smaller than 0,78% between the Albedo method and code ANISN. In relation to the Diffusion method were obtained errors smaller than 0,35%, showing the effectiveness of the Albedo method applied to criticality analysis. The easiness of application, simplicity and clarity of the Albedo method constitute a valuable instrument to neutronic calculations applied to nonmultiplying and multiplying media. (author)

An analytical approach for a nodal scheme of two-dimensional neutron transport problems

International Nuclear Information System (INIS)

Barichello, L.B.; Cabrera, L.C.; Prolo Filho, J.F.

2011-01-01

Research highlights: → Nodal equations for a two-dimensional neutron transport problem. → Analytical Discrete Ordinates Method. → Numerical results compared with the literature. - Abstract: In this work, a solution for a two-dimensional neutron transport problem, in cartesian geometry, is proposed, on the basis of nodal schemes. In this context, one-dimensional equations are generated by an integration process of the multidimensional problem. Here, the integration is performed for the whole domain such that no iterative procedure between nodes is needed. The ADO method is used to develop analytical discrete ordinates solution for the one-dimensional integrated equations, such that final solutions are analytical in terms of the spatial variables. The ADO approach along with a level symmetric quadrature scheme, lead to a significant order reduction of the associated eigenvalues problems. Relations between the averaged fluxes and the unknown fluxes at the boundary are introduced as the usually needed, in nodal schemes, auxiliary equations. Numerical results are presented and compared with test problems.

Thermal neutron moderating device

International Nuclear Information System (INIS)

Takigami, Hiroyuki.

1995-01-01

In a thermal neutron moderating device, superconductive coils for generating magnetic fields capable of applying magnetic fields vertical to the longitudinal direction of a thermal neutron passing tube, and superconductive coils for magnetic field gradient for causing magnetic field gradient in the longitudinal direction of the thermal neutron passing tube are disposed being stacked at the outside of the thermal neutron passing tube. When magnetic field gradient is present vertically to the direction of a magnetic moment, thermal neutrons undergo forces in the direction of the magnetic field gradient in proportion to the magnetic moment. Then, the magnetic moment of the thermal neutrons is aligned with the direction vertical to the passing direction of the thermal neutrons, to cause the magnetic field gradient in the passing direction of the thermal neutrons. The speed of the thermal neutrons can be optionally selected and the wavelength can freely be changed by applying forces to the thermal neutrons and changing the extent and direction of the magnetic field gradient. Superconductive coils are used as the coils for generating magnetic fields and the magnetic field gradient in order to change extremely high energy of the thermal neutrons. (N.H.)

Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

International Nuclear Information System (INIS)

Tzika, F.; Stamatelatos, I.E.

2004-01-01

Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample

Two-dimensional over-all neutronics analysis of the ITER device

Science.gov (United States)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

Two-dimensional over-all neutronics analysis of the ITER device

International Nuclear Information System (INIS)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi.

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR) and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li 2 O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No.5 and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design. (author)

Incorrectness of conventional one-dimensional parallel thermal resistance circuit model for two-dimensional circular composite pipes

International Nuclear Information System (INIS)

Wong, K.-L.; Hsien, T.-L.; Chen, W.-L.; Yu, S.-J.

2008-01-01

This study is to prove that two-dimensional steady state heat transfer problems of composite circular pipes cannot be appropriately solved by the conventional one-dimensional parallel thermal resistance circuits (PTRC) model because its interface temperatures are not unique. Thus, the PTRC model is definitely different from its conventional recognized analogy, parallel electrical resistance circuits (PERC) model, which has unique node electric voltages. Two typical composite circular pipe examples are solved by CFD software, and the numerical results are compared with those obtained by the PTRC model. This shows that the PTRC model generates large error. Thus, this conventional model, introduced in most heat transfer text books, cannot be applied to two-dimensional composite circular pipes. On the contrary, an alternative one-dimensional separately series thermal resistance circuit (SSTRC) model is proposed and applied to a two-dimensional composite circular pipe with isothermal boundaries, and acceptable results are returned

Fuel assembly inspection by three-dimensional neutron radiography

International Nuclear Information System (INIS)

Lapinski, N.P.; Reimann, K.J.; Berger, H.

1979-01-01

Radiographic inspection of complex objects such as fuel subassemblies often presents problems because superimposition of images at different depths in the object complicates interpretation. One method for obtaining and displaying three-dimensional neutron radiographic images in multiple-film laminagraphy; a series of radiographs generated at different angular orientations are superimposed to provide focussed images of any object plane. In the present work multiple-film neutron laminagraphs were generated using direct and indirect exposure techniques, with neutrons in thermal, epithermal, and fast energy ranges

TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

International Nuclear Information System (INIS)

Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman

1988-01-01

TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory

TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

Energy Technology Data Exchange (ETDEWEB)

Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang

1988-03-01

TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.

SNAP-3D: a three-dimensional neutron diffusion code

International Nuclear Information System (INIS)

McCallien, C.W.J.

1975-10-01

A preliminary report is presented describing the data requirements of a one- two- or three-dimensional multi-group diffusion code, SNAP-3D. This code is primarily intended for neutron diffusion calculations but it can also carry out gamma calculations if the diffuse approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. It is assumed the reader is familiar with the older, two-dimensional code SNAP and can refer to the report [TRG-Report-1990], describing it. The present report concentrates on the enhancements to SNAP that have been made to produce the three-dimensional version, SNAP-3D, and is intended to act a a guide on data preparation until a single, comprehensive report can be published. (author)

The inaccuracy of conventional one-dimensional parallel thermal resistance circuit model for two-dimensional composite walls

International Nuclear Information System (INIS)

Wong, K.-L.; Hsien, T.-L.; Hsiao, M.-C.; Chen, W.-L.; Lin, K.-C.

2008-01-01

This investigation is to show that two-dimensional steady state heat transfer problems of composite walls should not be solved by the conventionally one-dimensional parallel thermal resistance circuits (PTRC) model because the interface temperatures are not unique. Thus PTRC model cannot be used like its conventional recognized analogy, parallel electrical resistance circuits (PERC) model which has the unique node electric voltage. Two typical composite wall examples, solved by CFD software, are used to demonstrate the incorrectness. The numerical results are compared with those obtained by PTRC model, and very large differences are observed between their results. This proves that the application of conventional heat transfer PTRC model to two-dimensional composite walls, introduced in most heat transfer text book, is totally incorrect. An alternative one-dimensional separately series thermal resistance circuit (SSTRC) model is proposed and applied to the two-dimensional composite walls with isothermal boundaries. Results with acceptable accuracy can be obtained by the new model

Iterative Two- and One-Dimensional Methods for Three-Dimensional Neutron Diffusion Calculations

International Nuclear Information System (INIS)

Lee, Hyun Chul; Lee, Deokjung; Downar, Thomas J.

2005-01-01

Two methods are proposed for solving the three-dimensional neutron diffusion equation by iterating between solutions of the two-dimensional (2-D) radial and one-dimensional (1-D) axial solutions. In the first method, the 2-D/1-D equations are coupled using a current correction factor (CCF) with the average fluxes of the lower and upper planes and the axial net currents at the plane interfaces. In the second method, an analytic expression for the axial net currents at the interface of the planes is used for planar coupling. A comparison of the new methods is made with two previously proposed methods, which use interface net currents and partial currents for planar coupling. A Fourier convergence analysis of the four methods was performed, and results indicate that the two new methods have at least three advantages over the previous methods. First, the new methods are unconditionally stable, whereas the net current method diverges for small axial mesh size. Second, the new methods provide better convergence performance than the other methods in the range of practical mesh sizes. Third, the spectral radii of the new methods asymptotically approach zero as the mesh size increases, while the spectral radius of the partial current method approaches a nonzero value as the mesh size increases. Of the two new methods proposed here, the analytic method provides a smaller spectral radius than the CCF method, but the CCF method has several advantages over the analytic method in practical applications

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

International Nuclear Information System (INIS)

Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.

2007-01-01

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required

Neutronics code VALE for two-dimensional triagonal (hexagonal) and three-dimensional geometries

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.

1981-08-01

This report documents the computer code VALE designed to solve multigroup neutronics problems with the diffusion theory approximation to neutron transport for a triagonal arrangement of mesh points on planes in two- and three-dimensional geometry. This code parallels the VENTURE neutronics code in the local computation system, making exposure and fuel management capabilities available. It uses and generates interface data files adopted in the cooperative effort sponsored by Reactor Physics RRT Division of the US DOE. The programming in FORTRAN is straightforward, although data is transferred in blocks between auxiliary storage devices and main core, and direct access schemes are used. The size of problems which can be handled is essentially limited only by cost of calculation since the arrays are variably dimensioned. The memory requirement is held down while data transfer during iteration is increased only as necessary with problem size. There is provision for the more common boundary conditions including the repeating boundary, 180 0 rotational symmetry, and the rotational symmetry conditions for the 30 0 , 60 0 , and 120 0 triangular grids on planes. A variety of types of problems may be solved: the usual neutron flux eignevalue problem, or a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations. The adjoint problem and fixed source problem may be solved, as well as the dominating higher harmonic, or the importance problem for an arbitrary fixed source

Influence of an SN solver in a fine-mesh neutronics/thermal-hydraulics framework

International Nuclear Information System (INIS)

Jareteg, Klas; Vinai, Paolo; Demaziere, Christophe; Sasic, Srdjan

2015-01-01

In this paper a study on the influence of a neutron discrete ordinates (S N ) solver within a fine-mesh neutronic/thermal-hydraulic methodology is presented. The methodology consists of coupling a neutronic solver with a single-phase fluid solver, and it is aimed at computing the two fields on a three-dimensional (3D) sub-pin level. The cross-sections needed for the neutron transport equations are pre-generated using a Monte Carlo approach. The coupling is resolved in an iterative manner with full convergence of both fields. A conservative transfer of the full 3D information is achieved, allowing for a proper coupling between the neutronic and the thermal-hydraulic meshes on the finest calculated scales. The discrete ordinates solver is benchmarked against a Monte Carlo reference solution for a two-dimensional (2D) system. The results confirm the need of a high number of ordinates, giving a satisfactory accuracy in k eff and scalar flux profile applying S 16 for 16 energy groups. The coupled framework is used to compare the S N implementation and a solver based on the neutron diffusion approximation for a full 3D system of a quarter of a symmetric, 7x7 array in an infinite lattice setup. In this case, the impact of the discrete ordinates solver shows to be significant for the coupled system, as demonstrated in the calculations of the temperature distributions. (author)

Complex of two-dimensional multigroup programs for neutron-physical computations of nuclear reactor

International Nuclear Information System (INIS)

Karpov, V.A.; Protsenko, A.N.

1975-01-01

Briefly stated mathematical aspects of the two-dimensional multigroup method of neutron-physical computation of nuclear reactor. Problems of algorithmization and BESM-6 computer realisation of multigroup diffuse approximations in hexagonal and rectangular calculated lattices are analysed. The results of computation of fast critical assembly having complicated composition of the core are given. The estimation of computation accuracy of criticality, neutron fields distribution and efficiency of absorbing rods by means of computer programs developed is done. (author)

Monte Carlo simulations of the pulsed thermal neutron flux in two-region hydrogenous systems (using standard MCNP data libraries)

International Nuclear Information System (INIS)

Wiacek, U.; Krynicka, E.

2005-02-01

Monte Carlo simulations of the pulsed neutron experiment in two- region systems (two concentric spheres and two coaxial finite cylinders) are presented. The MCNP code is used. Aqueous solutions of H 3 BO 3 or KCl are used in the inner region. The outer region is the moderator of Plexiglas. Standard data libraries of the thermal neutron scattering cross-sections of hydrogen in hydrogenous substances are used. The time-dependent thermal neutron transport is simulated when the inner region has a constant size and the external size of the surrounding outer region is variable. The time decay constant of the thermal neutron flux in the system is found in each simulation. The results of the simulations are compared with results of real pulsed neutron experiments on the corresponding systems. (author)

Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

International Nuclear Information System (INIS)

Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

2005-01-01

Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

The Thermal Neutron Beam Option for NECTAR at MLZ

Science.gov (United States)

Mühlbauer, M. J.; Bücherl, T.; Genreith, C.; Knapp, M.; Schulz, M.; Söllradl, S.; Wagner, F. M.; Ehrenberg, H.

The beam port SR10 at the neutron source FRM II of Heinz Maier-Leibnitz Zentrum (MLZ) is equipped with a moveable assembly of two uranium plates, which can be placed in front of the entrance window of the beam tube via remote control. With these plates placed in their operating position the thermal neutron spectrum produced by the neutron source FRM II is converted to fission neutrons with 1.9 MeV of mean energy. This fission neutron spectrum is routinely used for medical applications at the irradiation facility MEDAPP, for neutron radiography and tomography experiments at the facility NECTAR and for materials testing. If, however, the uranium plates are in their stand-by position far off the tip of the beam tube and the so-called permanent filter for thermal neutrons is removed, thermal neutrons originating from the moderator tank enter the beam tube and a thermal spectrum becomes available for irradiation or activation of samples. By installing a temporary flight tube the beam may be used for thermal neutron radiography and tomography experiments at NECTAR. The thermal neutron beam option not only adds a pure thermal neutron spectrum to the energy ranges available for neutron imaging at MLZ instruments but it also is an unique possibility to combine two quite different neutron energy ranges at a single instrument including their respective advantages. The thermal neutron beam option for NECTAR is funded by BMBF in frame of research project 05K16VK3.

Comparison of preconditioned generalized conjugate gradient methods to two-dimensional neutron and photon transport equation

International Nuclear Information System (INIS)

Chen, G.S.

1997-01-01

We apply and compare the preconditioned generalized conjugate gradient methods to solve the linear system equation that arises in the two-dimensional neutron and photon transport equation in this paper. Several subroutines are developed on the basis of preconditioned generalized conjugate gradient methods for time-independent, two-dimensional neutron and photon transport equation in the transport theory. These generalized conjugate gradient methods are used. TFQMR (transpose free quasi-minimal residual algorithm), CGS (conjuage gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These sub-routines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. (author)

Matrix-type multiple reciprocity boundary element method for solving three-dimensional two-group neutron diffusion equations

International Nuclear Information System (INIS)

Itagaki, Masafumi; Sahashi, Naoki.

1997-01-01

The multiple reciprocity boundary element method has been applied to three-dimensional two-group neutron diffusion problems. A matrix-type boundary integral equation has been derived to solve the first and the second group neutron diffusion equations simultaneously. The matrix-type fundamental solutions used here satisfy the equation which has a point source term and is adjoint to the neutron diffusion equations. A multiple reciprocity method has been employed to transform the matrix-type domain integral related to the fission source into an equivalent boundary one. The higher order fundamental solutions required for this formulation are composed of a series of two types of analytic functions. The eigenvalue itself is also calculated using only boundary integrals. Three-dimensional test calculations indicate that the present method provides stable and accurate solutions for criticality problems. (author)

Solution and Study of the Two-Dimensional Nodal Neutron Transport Equation

International Nuclear Information System (INIS)

Panta Pazos, Ruben; Biasotto Hauser, Eliete; Tullio de Vilhena, Marco

2002-01-01

In the last decade Vilhena and coworkers reported an analytical solution to the two-dimensional nodal discrete-ordinates approximations of the neutron transport equation in a convex domain. The key feature of these works was the application of the combined collocation method of the angular variable and nodal approach in the spatial variables. By nodal approach we mean the transverse integration of the SN equations. This procedure leads to a set of one-dimensional S N equations for the average angular fluxes in the variables x and y. These equations were solved by the old version of the LTS N method, which consists in the application of the Laplace transform to the set of nodal S N equations and solution of the resulting linear system by symbolic computation. It is important to recall that this procedure allow us to increase N the order of S N up to 16. To overcome this drawback we step forward performing a spectral painstaking analysis of the nodal S N equations for N up to 16 and we begin the convergence of the S N nodal equations defining an error for the angular flux and estimating the error in terms of the truncation error of the quadrature approximations of the integral term. Furthermore, we compare numerical results of this approach with those of other techniques used to solve the two-dimensional discrete approximations of the neutron transport equation. (authors)

TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

International Nuclear Information System (INIS)

Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

1984-06-01

The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

Decay of the pulsed thermal neutron flux in two-zone hydrogenous systems - Monte Carlo simulations using MCNP standard data libraries

International Nuclear Information System (INIS)

Wiacek, Urszula; Krynicka, Ewa

2006-01-01

Pulsed neutron experiments in two-zone spherical and cylindrical geometry has been simulated using the MCNP code. The systems are built of hydrogenous materials. The inner zone is filled with aqueous solutions of absorbers (H 3 BO 3 or KCl). It is surrounded by the outer zone built of Plexiglas. The system is irradiated with the pulsed thermal neutron flux and the thermal neutron decay in time is observed. Standard data libraries of the thermal neutron scattering cross-sections of hydrogen in hydrogenous substances have been used to simulate the neutron transport. The time decay constant of the fundamental mode of the thermal neutron flux determined in each simulation has been compared with the corresponding result of the real pulsed neutron experiment

Development of a two-dimensional imaging detector based on a neutron scintillator with wavelength-shifting fibers

CERN Document Server

Sakai, K; Oku, T; Morimoto, K; Shimizu, H M; Tokanai, F; Gorin, A; Manuilov, I V; Ryazantsev, A; Ino, T; Kuroda, K; Suzuki, J

2002-01-01

For evaluating neutron optical devices, a two-dimensional (2D) detector based on a neutron scintillator with wavelength-shifting fibers has been developed at RIKEN. We have investigated a ZnS(Ag)+LiF and a Li glass plate as neutron scintillators with the coding technique for realizing the large sensitive area of 50 x 50 mm sup 2. After fabricating the 2D detector, its performance was tested using cold neutrons at JAERI. As a result, a spatial resolution of propor to 1.0 mm was obtained. (orig.)

Design of hyper-thermal neutron irradiation fields for neutron capture therapy in KUR-heavy water neutron irradiation facility. Mounting of hyper-thermal neutron converter in therapeutic collimator

International Nuclear Information System (INIS)

Sakurai, Y.; Kobayashi, T.

2001-01-01

Neutron capture therapy (NCP) using thermal neutron needs to improve of depth dose distribution in a living body. Epi-thermal neutron following moderation of fast neutron is usually used for improving of the depth dose distribution. The moderation method of fast neutron, however, gets mixed some of high energy neutron which give some of serious effects to a living body, and involves the difficulty for collimation of thermal neutron to the diseased part. Hyper-thermal neutrons, which are in an energy range of 0.1-3 eV at high temperature side of thermal neutron, are under consideration for application to the NCP. The hyper-thermal neutrons can be produced by up-scattering of thermal neutron in a high temperature material. Fast neutron components in collimator for the NCP reduce on application of the up-scattering method. Graphite at high temperature (>1000k) is used as a hyper-thermal neutron converter. The hyper-thermal neutron converter is planted to mount on therapeutic collimator which is located at the nearest side of patient for the NCP. Total neutron flux, ratio of hyper-thermal neutron to total neutron, and ratio of gamma-ray dose to neutron flux are calculated as a function of thickness of the graphite converter using monte carlo code MCNP-V4B. (M. Suetake)

Hyper-thermal neutron irradiation field for neutron capture therapy

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1994-01-01

The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwell distribution higher than the room temperature of 300 K, has been studied in order to improve the thermal neutron flux distribution in a living body for a deep-seated tumor in neutron capture therapy (NCT). Simulation calculations using MCNP-V3 were carried out in order to investigate the characteristics of the hyper-thermal neutron irradiation field. From the results of simulation calculations, the following were confirmed: (i) The irradiation field of the hyper-thermal neutrons is feasible by using some scattering materials with high temperature, such as Be, BeO, C, SiC and ZrH 1.7 . Especially, ZrH 1.7 is thought to be the best material because of good characteristics of up-scattering for thermal neutrons. (ii) The ZrH 1.7 of 1200 K yields the hyper-thermal neutrons of a Maxwell-like distribution at about 2000 K and the treatable depth is about 1.5 cm larger comparing with the irradiation of the thermal neutrons of 300 K. (iii) The contamination by the secondary gamma-rays from the scattering materials can be sufficiently eliminated to the tolerance level for NCT through the bismuth layer, without the larger change of the energy spectrum of hyper-thermal neutrons. ((orig.))

Test of quantum thermalization in the two-dimensional transverse-field Ising model.

Science.gov (United States)

Blaß, Benjamin; Rieger, Heiko

2016-12-01

We study the quantum relaxation of the two-dimensional transverse-field Ising model after global quenches with a real-time variational Monte Carlo method and address the question whether this non-integrable, two-dimensional system thermalizes or not. We consider both interaction quenches in the paramagnetic phase and field quenches in the ferromagnetic phase and compare the time-averaged probability distributions of non-conserved quantities like magnetization and correlation functions to the thermal distributions according to the canonical Gibbs ensemble obtained with quantum Monte Carlo simulations at temperatures defined by the excess energy in the system. We find that the occurrence of thermalization crucially depends on the quench parameters: While after the interaction quenches in the paramagnetic phase thermalization can be observed, our results for the field quenches in the ferromagnetic phase show clear deviations from the thermal system. These deviations increase with the quench strength and become especially clear comparing the shape of the thermal and the time-averaged distributions, the latter ones indicating that the system does not completely lose the memory of its initial state even for strong quenches. We discuss our results with respect to a recently formulated theorem on generalized thermalization in quantum systems.

Two-dimensional fruit ripeness estimation using thermal imaging

Science.gov (United States)

Sumriddetchkajorn, Sarun; Intaravanne, Yuttana

2013-06-01

Some green fruits do not change their color from green to yellow when being ripe. As a result, ripeness estimation via color and fluorescent analytical approaches cannot be applied. In this article, we propose and show for the first time how a thermal imaging camera can be used to two-dimensionally classify fruits into different ripeness levels. Our key idea relies on the fact that the mature fruits have higher heat capacity than the immature ones and therefore the change in surface temperature overtime is slower. Our experimental proof of concept using a thermal imaging camera shows a promising result in non-destructively identifying three different ripeness levels of mangoes Mangifera indica L.

A two-dimensional, semi-analytic expansion method for nodal calculations

International Nuclear Information System (INIS)

Palmtag, S.P.

1995-08-01

Most modern nodal methods used today are based upon the transverse integration procedure in which the multi-dimensional flux shape is integrated over the transverse directions in order to produce a set of coupled one-dimensional flux shapes. The one-dimensional flux shapes are then solved either analytically or by representing the flux shape by a finite polynomial expansion. While these methods have been verified for most light-water reactor applications, they have been found to have difficulty predicting the large thermal flux gradients near the interfaces of highly-enriched MOX fuel assemblies. A new method is presented here in which the neutron flux is represented by a non-seperable, two-dimensional, semi-analytic flux expansion. The main features of this method are (1) the leakage terms from the node are modeled explicitly and therefore, the transverse integration procedure is not used, (2) the corner point flux values for each node are directly edited from the solution method, and a corner-point interpolation is not needed in the flux reconstruction, (3) the thermal flux expansion contains hyperbolic terms representing analytic solutions to the thermal flux diffusion equation, and (4) the thermal flux expansion contains a thermal to fast flux ratio term which reduces the number of polynomial expansion functions needed to represent the thermal flux. This new nodal method has been incorporated into the computer code COLOR2G and has been used to solve a two-dimensional, two-group colorset problem containing uranium and highly-enriched MOX fuel assemblies. The results from this calculation are compared to the results found using a code based on the traditional transverse integration procedure

Simultaneous thermal neutron decay time and porosity logging system

International Nuclear Information System (INIS)

Smith, H.D. Jr.; Smith, M.P.; Schultz, W.E.

1979-01-01

A simultaneous pulsed neutron porosity and thermal neutron capture cross section logging system is provided for radiological well logging of subsurface earth formations. A logging tool provided with a 14 MeV pulsed neutron source, an epithermal neutron detector, and a combination gamma ray and fast neutron detector is moved through a borehole. Repetitive bursts of neutrons irradiate the earth formations; and, during the bursts, the fast neutron and epithermal neutron populations are sampled. During the interval between bursts the thermal neutron capture gamma ray population is sampled in two or more time intervals. The fast and epithermal neutron population measurements are combined to provide a measurement of formation porosity phi. The capture gamma ray measurements are combined to provide a simultaneous determination of the thermal neutron capture cross section Σ

STEADY-SHIP: a computer code for three-dimensional nuclear and thermal-hydraulic analyses of marine reactors

International Nuclear Information System (INIS)

Itagaki, Masafumi; Naito, Yoshitaka; Tokuno, Yukio; Matsui, Yasushi.

1988-01-01

A code STEADY-SHIP has been developed to calculate three-dimensional distributions of neutron flux, power and coolant temperature in the reactor core of the nuclear ship MUTSU. The code consists of two parts, that is, a few-group three-dimensional neutron diffusion module DIFFUSION-SHIP and a thermal-hydraulic module HYDRO-SHIP: In the DIFFUSION-SHIP the leakage iteration method is used for solving the three-dimensional neutron diffusion equation with small computer core memory and short computing time; The HYDRO-SHIP performs the general thermal-hydraulic calculation for evaluating feedbacks required in the neutronic calculation by the DIFFUSION-SHIP. The macroscopic nuclear constants are generated by a module CROSS-SHIP as functions of xenon poison, fuel temperature, moderator temperature and moderator density. A module LOCAL-FINE has the capability of computing a detailed rod power distribution for each local node in the core, using the boundary conditions on the surface of the node which were supplied by the STEADY-SHIP whole-core calculation. The applicability of this code to marine reactors has been demonstrated by comparing the computed results with the data measured during the MUTSU land-loaded core critical experiments and with the data obtained during the hot-zero-power tests performed for the actual MUTSU plant. (author)

Test of quantum thermalization in the two-dimensional transverse-field Ising model

Science.gov (United States)

Blaß, Benjamin; Rieger, Heiko

2016-01-01

We study the quantum relaxation of the two-dimensional transverse-field Ising model after global quenches with a real-time variational Monte Carlo method and address the question whether this non-integrable, two-dimensional system thermalizes or not. We consider both interaction quenches in the paramagnetic phase and field quenches in the ferromagnetic phase and compare the time-averaged probability distributions of non-conserved quantities like magnetization and correlation functions to the thermal distributions according to the canonical Gibbs ensemble obtained with quantum Monte Carlo simulations at temperatures defined by the excess energy in the system. We find that the occurrence of thermalization crucially depends on the quench parameters: While after the interaction quenches in the paramagnetic phase thermalization can be observed, our results for the field quenches in the ferromagnetic phase show clear deviations from the thermal system. These deviations increase with the quench strength and become especially clear comparing the shape of the thermal and the time-averaged distributions, the latter ones indicating that the system does not completely lose the memory of its initial state even for strong quenches. We discuss our results with respect to a recently formulated theorem on generalized thermalization in quantum systems. PMID:27905523

Improvement of neutron kinetics module in TRAC-BF1code: one-dimensional nodal collocation method

Energy Technology Data Exchange (ETDEWEB)

Jambrina, Ana; Barrachina, Teresa; Miro, Rafael; Verdu, Gumersindo, E-mail: ajambrina@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidade Politecnica de Valencia (UPV), Valencia (Spain); Soler, Amparo, E-mail: asoler@iberdrola.es [SEA Propulsion S.L., Madrid (Spain); Concejal, Alberto, E-mail: acbe@iberdrola.es [Iberdrola Ingenieria y Construcion S.A.U., Madrid (Spain)

2013-07-01

The TRAC-BF1 one-dimensional kinetic model is a formulation of the neutron diffusion equation in the two energy groups' approximation, based on the analytical nodal method (ANM). The advantage compared with a zero-dimensional kinetic model is that the axial power profile may vary with time due to thermal-hydraulic parameter changes and/or actions of the control systems but at has the disadvantages that in unusual situations it fails to converge. The nodal collocation method developed for the neutron diffusion equation and applied to the kinetics resolution of TRAC-BF1 thermal-hydraulics, is an adaptation of the traditional collocation methods for the discretization of partial differential equations, based on the development of the solution as a linear combination of analytical functions. It has chosen to use a nodal collocation method based on a development of Legendre polynomials of neutron fluxes in each cell. The qualification is carried out by the analysis of the turbine trip transient from the NEA benchmark in Peach Bottom NPP using both the original 1D kinetics implemented in TRAC-BF1 and the 1D nodal collocation method. (author)

A two-dimensional detector with delay line readout for slow neutron fields measurements

International Nuclear Information System (INIS)

Cheremukhina, G.A.; Chernenko, S.P.; Ivanov, A.B.

1992-01-01

This article presents the description of a two-dimensional detector of slow neutrons together with its readout and data acquisition electronics based on a PC/AT> The detector with a sensitive area of 260x140 mm 2 is based on a high pressure multiwire proportional chamber with delay line readout and gas filling of 3.0 atm. 3 He + propane. 25 refs.; 10 figs.; 2 tabs

Pulsed thermal neutron source at the fast neutron generator.

Science.gov (United States)

Tracz, Grzegorz; Drozdowicz, Krzysztof; Gabańska, Barbara; Krynicka, Ewa

2009-06-01

A small pulsed thermal neutron source has been designed based on results of the MCNP simulations of the thermalization of 14 MeV neutrons in a cluster-moderator which consists of small moderating cells decoupled by an absorber. Optimum dimensions of the single cell and of the whole cluster have been selected, considering the thermal neutron intensity and the short decay time of the thermal neutron flux. The source has been built and the test experiments have been performed. To ensure the response is not due to the choice of target for the experiments, calculations have been done to demonstrate the response is valid regardless of the thermalization properties of the target.

Fail-safe neutron shutter used for thermal neutron radiography

International Nuclear Information System (INIS)

Sachs, R.D.; Morris, R.A.

1976-11-01

A fail-safe, reliable, easy-to-use neutron shutter was designed, built, and put into operation at the Omega West Reactor, Los Alamos Scientific Laboratory. The neutron shutter will be used primarily to perform thermal neutron radiography, but is also available for a highly collimated source of thermal neutrons [neutron flux = 3.876 x 10 6 (neutrons)/(cm 2 .s)]. Neutron collimator sizes of either 10.16 by 10.16 cm or 10.16 by 30.48 cm are available

Solution of two-dimensional neutron diffusion equation for triangular region by finite Fourier transformation

International Nuclear Information System (INIS)

Kobayashi, Keisuke; Ishibashi, Hideo

1978-01-01

A two-dimensional neutron diffusion equation for a triangular region is shown to be solved by the finite Fourier transformation. An application of the Fourier transformation to the diffusion equation for triangular region yields equations whose unknowns are the expansion coefficients of the neutron flux and current in Fourier series or Legendre polynomials expansions only at the region boundary. Some numerical calculations have revealed that the present technique gives accurate results. It is shown also that the solution using the expansion in Legendre polynomials converges with relatively few terms even if the solution in Fourier series exhibits the Gibbs' phenomenon. (auth.)

Two-dimensional neutron scattering in a floating heavy water bridge

International Nuclear Information System (INIS)

Fuchs, Elmar C; Baroni, Patrick; Noirez, Laurence; Bitschnau, Brigitte

2010-01-01

When a high voltage is applied to pure water in two filled beakers kept close to each other, a connection forms spontaneously, giving the impression of a floating water bridge. This phenomenon is of special interest, since it comprises a number of phenomena currently tackled in modern water science. In this work, the first two-dimensional structural study of a floating heavy water bridge is presented as a function of the azimuthal angle. A small anisotropy in the angular distribution of the intensity of the first structural peak was observed, indicating a preferred orientation of a part of the D 2 O molecules along the electric field lines without breaking the local tetrahedral symmetry. The experiment is carried out by neutron scattering on a D 2 O bridge.

Two-dimensional neutron scattering in a floating heavy water bridge

Science.gov (United States)

Fuchs, Elmar C.; Baroni, Patrick; Bitschnau, Brigitte; Noirez, Laurence

2010-03-01

When a high voltage is applied to pure water in two filled beakers kept close to each other, a connection forms spontaneously, giving the impression of a floating water bridge. This phenomenon is of special interest, since it comprises a number of phenomena currently tackled in modern water science. In this work, the first two-dimensional structural study of a floating heavy water bridge is presented as a function of the azimuthal angle. A small anisotropy in the angular distribution of the intensity of the first structural peak was observed, indicating a preferred orientation of a part of the D2O molecules along the electric field lines without breaking the local tetrahedral symmetry. The experiment is carried out by neutron scattering on a D2O bridge.

Non-destructive studies of fuel pellets by neutron resonance absorption radiography and thermal neutron radiography

Energy Technology Data Exchange (ETDEWEB)

Tremsin, A.S., E-mail: ast@ssl.berkeley.edu [University of California, Berkeley, CA 94720 (United States); Vogel, S.C.; Mocko, M.; Bourke, M.A.M.; Yuan, V.; Nelson, R.O.; Brown, D.W. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Feller, W.B. [NOVA Scientific, Inc., 10 Picker Rd., Sturbridge, MA 01566 (United States)

2013-09-15

Many isotopes in nuclear materials exhibit strong peaks in neutron absorption cross sections in the epithermal energy range (1–1000 eV). These peaks (often referred to as resonances) occur at energies specific to particular isotopes, providing a means of isotope identification and concentration measurements. The high penetration of epithermal neutrons through most materials is very useful for studies where samples consist of heavy-Z elements opaque to X-rays and sometimes to thermal neutrons as well. The characterization of nuclear fuel elements in their cladding can benefit from the development of high resolution neutron resonance absorption imaging (NRAI), enabled by recently developed spatially-resolved neutron time-of-flight detectors. In this technique the neutron transmission of the sample is measured as a function of spatial location and of neutron energy. In the region of the spectra that borders the resonance energy for a particular isotope, the reduction in transmission can be used to acquire an image revealing the 2-dimensional distribution of that isotope within the sample. Provided that the energy of each transmitted neutron is measured by the neutron detector used and the irradiated sample possesses neutron absorption resonances, then isotope-specific location maps can be acquired simultaneously for several isotopes. This can be done even in the case where samples are opaque or have very similar transmission for thermal neutrons and X-rays or where only low concentrations of particular isotopes are present (<0.1 atom% in some cases). Ultimately, such radiographs of isotope location can be utilized to measure isotope concentration, and can even be combined to produce three-dimensional distributions using tomographic methods. In this paper we present the proof-of-principle of NRAI and transmission Bragg edge imaging performed at Flight Path 5 (FP5) at the LANSCE pulsed, moderated neutron source of Los Alamos National Laboratory. A set of urania mockup

Feasibility study on using imaging plates to estimate thermal neutron fluence in neutron-gamma mixed fields

International Nuclear Information System (INIS)

Fujibuchi, T.; Tanabe, Y.; Sakae, T.; Terunuma, T.; Isobe, T.; Kawamura, H.; Yasuoka, K.; Matsumoto, T.; Harano, H.; Nishiyama, J.; Masuda, A.; Nohtomi, A.

2011-01-01

In current radiotherapy, neutrons are produced in a photonuclear reaction when incident photon energy is higher than the threshold. In the present study, a method of discriminating the neutron component was investigated using an imaging plate (IP) in the neutron-gamma-ray mixed field. Two types of IP were used: a conventional IP for beta- and gamma rays, and an IP doped with Gd for detecting neutrons. IPs were irradiated in the mixed field, and the photo-stimulated luminescence (PSL) intensity of the thermal neutron component was discriminated using an expression proposed herein. The PSL intensity of the thermal neutron component was proportional to thermal neutron fluence. When additional irradiation of photons was added to constant neutron irradiation, the PSL intensity of the thermal neutron component was not affected. The uncertainty of PSL intensities was approximately 11.4 %. This method provides a simple and effective means of discriminating the neutron component in a mixed field. (authors)

Experimental characterization of semiconductor-based thermal neutron detectors

Energy Technology Data Exchange (ETDEWEB)

Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Bortot, D.; Pola, A.; Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, via La Masa 34, 20156 Milano (Italy); INFN—Milano, Via Celoria 16, 20133 Milano (Italy); Gómez-Ros, J.M. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain); Sacco, D. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); INAIL—DIT, Via di Fontana Candida 1, 00040 Monteporzio Catone (Italy); Esposito, A.; Gentile, A.; Buonomo, B. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Palomba, M.; Grossi, A. [ENEA Triga RC-1C.R. Casaccia, via Anguillarese 301, 00060 S. Maria di Galeria, Roma (Italy)

2015-04-21

In the framework of NESCOFI@BTF and NEURAPID projects, active thermal neutron detectors were manufactured by depositing appropriate thickness of {sup 6}LiF on commercially available windowless p–i–n diodes. Detectors with different radiator thickness, ranging from 5 to 62 μm, were manufactured by evaporation-based deposition technique and exposed to known values of thermal neutron fluence in two thermal neutron facilities exhibiting different irradiation geometries. The following properties of the detector response were investigated and presented in this work: thickness dependence, impact of parasitic effects (photons and epithermal neutrons), linearity, isotropy, and radiation damage following exposure to large fluence (in the order of 10{sup 12} cm{sup −2})

One-dimensional neutron imager for the Sandia Z facility.

Science.gov (United States)

Fittinghoff, David N; Bower, Dan E; Hollaway, James R; Jacoby, Barry A; Weiss, Paul B; Buckles, Robert A; Sammons, Timothy J; McPherson, Leroy A; Ruiz, Carlos L; Chandler, Gordon A; Torres, José A; Leeper, Ramon J; Cooper, Gary W; Nelson, Alan J

2008-10-01

A multiinstitution collaboration is developing a neutron imaging system for the Sandia Z facility. The initial system design is for slit aperture imaging system capable of obtaining a one-dimensional image of a 2.45 MeV source producing 5x10(12) neutrons with a resolution of 320 microm along the axial dimension of the plasma, but the design being developed can be modified for two-dimensional imaging and imaging of DT neutrons with other resolutions. This system will allow us to understand the spatial production of neutrons in the plasmas produced at the Z facility.

One-dimensional nodal neutronics routines for the TRAC-BD1 thermal-hydraulics program

International Nuclear Information System (INIS)

Nigg, D.W.

1983-09-01

Nuclear reactor core transient neutronic behavior is currently modeled in the TRAC-BD1 code using a point-reactor kinetics formulation. This report describes a set of subroutines based on the Analytic Nodal Method that were written to provide TRAC-BD1 with a one-dimensional space-dependent neutronics capability. Use of the routines is illustrated with several test problems. The results of these problems show that the Analytic Nodal neutronics routines have desirable accuracy and computing time characteristics and should be a useful addition to TRAC-BD1

A study on the utilization of hyper-thermal neutrons for neutron capture therapy

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

1993-01-01

The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwellian distribution of a higher temperature than the room temperature of 300 K, was studied in order to improve the thermal neutron flux distribution at the deeper part in a living body for neutron capture therapy. Simulation calculations were carried out using MCNP-V3 in order to confirm the characteristics of hyper-thermal neutrons, i.e., (1) depth dependence of neutron energy spectrum, and (2) depth distribution of the reaction rate in a water phantom for materials with 1/v neutron absorption. It is confirmed that the hyper-thermal neutron irradiation can improve the thermal neutron flux distribution in the deeper and wider area in a living body compared with the thermal neutron irradiation. Practically, by the incidence of the hyper-thermal neutrons with a 3000 K Maxwellian distribution, the thermal neutron flux at 5 cm depth can be given about four times larger than by the incidence of the thermal neutrons of 300 K. (author)

Method and apparatus for measuring thermal neutron characteristics

International Nuclear Information System (INIS)

Johnstone, C.W.

1983-01-01

The thermal neutron decay characteristics of an earth formation are measured by detecting indications of the thermal neutron concentration in the formation during a selected set of two measurement intervals following irradiation of the formation with a burst of fast neutrons. These measurement intervals may comprise a sequence of time gates following a delay after the neutron burst. The duration of the neutron bursts, of the delay between the burst and the start of the sequence, and of the individual time gates, may all be adjusted by a common, selected one of a finite number of scale factor values. The set of two measurement intervals is selected from among a number of possible sets as a function of a previously measured value of the decay characteristic. Each measurement interval set is used over only a specific range of decay characteristic values for which it has been determined, in accordance with a previously established relationship between the decay characteristic value and a function of the thermal neutron concentration measurements for the set, to afford enhanced statistical accuracy in the measured value of the decay characteristic. (author)

Development of high flux thermal neutron generator for neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Vainionpaa, Jaakko H., E-mail: hannes@adelphitech.com [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Chen, Allan X.; Piestrup, Melvin A.; Gary, Charles K. [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Jones, Glenn [G& J Jones Enterprice, 7486 Brighton Ct, Dublin, CA 94568 (United States); Pantell, Richard H. [Department of Electrical Engineering, Stanford University, Stanford, CA (United States)

2015-05-01

The new model DD110MB neutron generator from Adelphi Technology produces thermal (<0.5 eV) neutron flux that is normally achieved in a nuclear reactor or larger accelerator based systems. Thermal neutron fluxes of 3–5 · 10{sup 7} n/cm{sup 2}/s are measured. This flux is achieved using four ion beams arranged concentrically around a target chamber containing a compact moderator with a central sample cylinder. Fast neutron yield of ∼2 · 10{sup 10} n/s is created at the titanium surface of the target chamber. The thickness and material of the moderator is selected to maximize the thermal neutron flux at the center. The 2.5 MeV neutrons are quickly thermalized to energies below 0.5 eV and concentrated at the sample cylinder. The maximum flux of thermal neutrons at the target is achieved when approximately half of the neutrons at the sample area are thermalized. In this paper we present simulation results used to characterize performance of the neutron generator. The neutron flux can be used for neutron activation analysis (NAA) prompt gamma neutron activation analysis (PGNAA) for determining the concentrations of elements in many materials. Another envisioned use of the generator is production of radioactive isotopes. DD110MB is small enough for modest-sized laboratories and universities. Compared to nuclear reactors the DD110MB produces comparable thermal flux but provides reduced administrative and safety requirements and it can be run in pulsed mode, which is beneficial in many neutron activation techniques.

Evaluation of two thermal neutron detection units consisting of ZnS/{sup 6}LiF scintillating layers with embedded WLS fibers read out with a SiPM

Energy Technology Data Exchange (ETDEWEB)

Mosset, J.-B., E-mail: jean-baptiste.mosset@a3.epfl.ch [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Stoykov, A.; Greuter, U.; Hildebrandt, M.; Schlumpf, N. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Van Swygenhoven, H. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Fédérale de Lausanne, CH-1015 Lausanne (Switzerland)

2014-11-11

Two single channel detection units for thermal neutron detection are investigated in a neutron beam. They consist of two ZnS/{sup 6}LiF scintillating layers sandwiching an array of WLS fibers. The pattern of these units can be repeated laterally and vertically in order to build up a one-dimensional position sensitive multi-channel detector with the needed sensitive surface and with the required neutron absorption probability. The originality of this work arises from the fact that the WLS fibers are read out with SiPMs instead of the traditionally used PMTs or MaPMTs. The signal processing system is based on a photon counting approach. For SiPMs with a dark count rate as high as 0.7 MHz, a trigger efficiency of 80% is achieved together with a system background rate lower than 10{sup −3}Hz and a dead time of 30μs. No change of performance is observed for neutron count rates of up to 3.6 kHz.

Thermal hydraulic and neutronic interaction in the rotating bed reactor

International Nuclear Information System (INIS)

Lee, C.C.

1986-01-01

Power transient characteristics in a rotating fluidized bed reactor (RBR) are investigated theoretically. A propellant flow perturbation is assumed to occur in an initially equilibrium state of the core. Transfer functions representing quasi-one-dimensional mutual feedback between thermal hydraulics and neutronics are developed and analyzed in the frequency domain. Neutronic responses are determined by Fermi-age theory for slowing down of fast neutrons and diffusion theory for thermal neutron distribution. Neutron leakage through the exhaust nozzle is accounted for by applying diffuse view factors similar to those applied in radiative heat transfer. The bed expansion behavior is described by a kinematic wave equation derived from the continuity of the gas phase. The drift flux approach is used to determine the yield fractions in the equilibrium bed. Thermal responses of fuel are evaluated by dividing it into several volume-averaged zones to better account for the transient effects over single zone models. Sample calculations are undertaken for the various operation conditions and design parameters of the RBR based on 250 MW/sub t/, 1000 MW/sub t/, and 5000 MW/sub t/ power reactors. The results show that power transients are dependent on the parametric changes of optical thickness and view factors

Transmutation of Minor Actinide in well thermalized neutron field and application of advanced neutron source (ANS)

International Nuclear Information System (INIS)

Iwasaki, Tomohiko; Hirakawa, Naohiro

1995-01-01

Transmutation of Minor Actinide (MA) in a well thermalized neutron field was studied. Since MA nuclides have large effective cross sections in the well thermalized neutron field, the transmutation in the well thermalized neutron field has an advantage of high transmutation rate. However, the transmutation rate largely decreases by accumulation of 246 Cm when MA is transmuted only in the well thermalized neutron field for a long period. An acceleration method of burn-up of 246 Cm was studied. High transmutation rate can be obtained by providing a neutron field with high flux in the energy region between 1 and 100 eV. Two stage transmutation using the well thermalized neutron field and this field can transmute MA rapidly. The applicability of the Advanced Neutron Source (ANS) to the transmutation of MA was examined for a typical MA with the composition in the high-level waste generated in the conventional PWR. If the ANS is applied without changing the fuel inventory, the amount of MA which corresponds to that produced by a conventional 1,175 MWe PWR in one year can be transmuted by the ANS in one year. Furthermore, the amount of the residual can be reduced to about 1g (10 -5 of the initial MA weight) by continuing the transmutation for 5 years owing to the two stage transmutation. (author)

Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

International Nuclear Information System (INIS)

Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.

2010-10-01

Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

Thermal neutron polarisation

International Nuclear Information System (INIS)

Satya Murthy, N.S.; Madhava Rao, L.

1984-01-01

The basic principle for the production of polarised thermal neutrons is discussed and the choice of various crystal monochromators surveyed. Brief mention of broad-spectrum polarisers is made. The application of polarised neutrons to the study of magnetisation density distributions in magnetic crystals, the dynamic concept of polarisation, principle and use of polarisation analysis, the neutron spin-echo technique are discussed. (author)

Study on the dose distribution of the mixed field with thermal and epi-thermal neutrons for neutron capture therapy

International Nuclear Information System (INIS)

Kobayashi, Tooru; Sakurai, Yoshinori; Kanda, Keiji

1994-01-01

Simulation calculations using DOT 3.5 were carried out in order to confirm the characteristics of depth-dependent dose distribution in water phantom dependent on incident neutron energy. The epithermal neutrons mixed to thermal neutron field is effective improving the thermal neutron depth-dose distribution for neutron capture therapy. A feasibility study on the neutron energy spectrum shifter was performed using ANISN-JR for the KUR Heavy Water Facility. The design of the neutron spectrum shifter is feasible, without reducing the performance as a thermal neutron irradiation field. (author)

Development of real time personal neutron dosimeter with two silicon detectors

Energy Technology Data Exchange (ETDEWEB)

Nakamura, T.; Tsujimura, N. [Tohoku Univ., Cyclotron and Radioisotope Center, Aoba, Aramaki, Aoba-ku (Japan); Yamano, T. [Tokyo Factory, Fuji Electric Co. Ltd., Tokyo (Japan)

1992-07-01

We developed a real time personal neutron dosimeter by using two types of silicon p-n junction detectors, thermal neutron sensor and fast neutron sensor. The thermal neutron sensor which is {sup 10}B doped n-type silicon with a polyethylene radiator mainly counts neutrons of energy front thermal to I MeV, and the fast neutron sensor which is p-type silicon with a polyethylene radiator is sensitive to neutrons above I MeV. The neutron sensitivity measurements revealed that the dosimeter has a rather flat response for dose equivalent from thermal to 15 MeV, excluding a drop from 50 keV to I MeV. In order to get conversion factor from counts to dose equivalent as accurately as possible, we performed the field test of the dosimeter calibration in several neutron-generating fields. By introducing the two-group dose estimation method, this dosimeter can give the neutron dose equivalent within about 50% errors. (author)

Low dimensional neutron moderators for enhanced source brightness

DEFF Research Database (Denmark)

Mezei, Ferenc; Zanini, Luca; Takibayev, Alan

2014-01-01

In a recent numerical optimization study we have found that liquid para-hydrogen coupled cold neutron moderators deliver 3–5 times higher cold neutron brightness at a spallation neutron source if they take the form of a flat, quasi 2-dimensional disc, in contrast to the conventional more voluminous...... for cold neutrons. This model leads to the conclusions that the optimal shape for high brightness para-hydrogen neutron moderators is the quasi 1-dimensional tube and these low dimensional moderators can also deliver much enhanced cold neutron brightness in fission reactor neutron sources, compared...... to the much more voluminous liquid D2 or H2 moderators currently used. Neutronic simulation calculations confirm both of these theoretical conclusions....

Thermal transport in a two-dimensional Z2 spin liquid

Science.gov (United States)

Metavitsiadis, Alexandros; Pidatella, Angelo; Brenig, Wolfram

2017-11-01

We study the dynamical thermal conductivity of the two-dimensional Kitaev spin model on the honeycomb lattice. We find a strongly temperature dependent low-frequency spectral intensity as a direct consequence of fractionalization of spins into mobile Majorana matter and a static Z2 gauge field. The latter acts as an emergent thermally activated disorder, leading to the appearance of a pseudogap which closes in the thermodynamic limit, indicating a dissipative heat conductor. Our analysis is based on complementary calculations of the current correlation function, comprising exact diagonalization by means of a complete summation over all gauge sectors, as well as a phenomenological mean-field treatment of thermal gauge fluctuations, valid at intermediate and high temperatures. The results will also be contrasted against the conductivity discarding gauge fluctuations.

Surface harmonics method for two-dimensional time-dependent neutron transport problems of square-lattice nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Boyarinov, V. F.; Kondrushin, A. E.; Fomichenko, P. A. [National Research Centre Kurchatov Institute, Kurchatov Sq. 1, Moscow (Russian Federation)

2013-07-01

Time-dependent equations of the Surface Harmonics Method (SHM) have been derived from the time-dependent neutron transport equation with explicit representation of delayed neutrons for solving the two-dimensional time-dependent problems. These equations have been realized in the SUHAM-TD code. The TWIGL benchmark problem has been used for verification of the SUHAM-TD code. The results of the study showed that computational costs required to achieve necessary accuracy of the solution can be an order of magnitude less than with the use of the conventional finite difference method (FDM). (authors)

Ultrafast carrier thermalization in lead iodide perovskite probed with two-dimensional electronic spectroscopy.

Science.gov (United States)

Richter, Johannes M; Branchi, Federico; Valduga de Almeida Camargo, Franco; Zhao, Baodan; Friend, Richard H; Cerullo, Giulio; Deschler, Felix

2017-08-29

In band-like semiconductors, charge carriers form a thermal energy distribution rapidly after optical excitation. In hybrid perovskites, the cooling of such thermal carrier distributions occurs on timescales of about 300 fs via carrier-phonon scattering. However, the initial build-up of the thermal distribution proved difficult to resolve with pump-probe techniques due to the requirement of high resolution, both in time and pump energy. Here, we use two-dimensional electronic spectroscopy with sub-10 fs resolution to directly observe the carrier interactions that lead to a thermal carrier distribution. We find that thermalization occurs dominantly via carrier-carrier scattering under the investigated fluences and report the dependence of carrier scattering rates on excess energy and carrier density. We extract characteristic carrier thermalization times from below 10 to 85 fs. These values allow for mobilities of 500 cm 2  V -1  s -1 at carrier densities lower than 2 × 10 19  cm -3 and limit the time for carrier extraction in hot carrier solar cells.Carrier-carrier scattering rates determine the fundamental limits of carrier transport and electronic coherence. Using two-dimensional electronic spectroscopy with sub-10 fs resolution, Richter and Branchi et al. extract carrier thermalization times of 10 to 85 fs in hybrid perovskites.

Studsvik thermal neutron facility

International Nuclear Information System (INIS)

Pettersson, O.A.; Larsson, B.; Grusell, E.; Svensson, P.

1992-01-01

The Studsvik thermal neutron facility at the R2-0 reactor originally designed for neutron capture radiography has been modified to permit irradiation of living cells and animals. A hole was drilled in the concrete shielding to provide a cylindrical channel with diameter of 25.3 cm. A shielding water tank serves as an entry holder for cells and animals. The advantage of this modification is that cells and animals can be irradiated at a constant thermal neutron fluence rate of approximately 10 9 n cm -2 s -1 (at 100 kW) without stopping and restarting the reactor. Topographic analysis of boron done by neutron capture autoradiography (NCR) can be irradiated under the same conditions as previously

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.H.

1992-09-01

Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by NRC regulations in 10CFR Part 71, the cask is subjected to an 800 degree C environment for 30 minutes. The neutron shield need not perform any shielding function during or after the thermal accident, but its behavior must not compromise the ability of the cask to contain the radioactive contents. In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-AL 9897, R. H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series, a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280 degree F. The neutron shield materials tested were boronated (0.8--4.5%) polymers (polypropylene, HDPE, NS-4). The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found

The discrete cones method for two-dimensional neutron transport calculations

International Nuclear Information System (INIS)

Watanabe, Y.; Maynard, C.W.

1986-01-01

A novel method, the discrete cones method (DC/sub N/), is proposed as an alternative to the discrete ordinates method (S/sub N/) for solutions of the two-dimensional neutron transport equation. The new method utilizes a new concept, discrete cones, which are made by partitioning a unit spherical surface that the direction vector of particles covers. In this method particles in a cone are simultaneously traced instead of those in discrete directions so that an anomaly of the S/sub N/ method, the ray effects, can be eliminated. The DC/sub N/ method has been formulated for X-Y geometry and a program has been creaed by modifying the standard S/sub N/ program TWOTRAN-II. Our sample calculations demonstrate a strong mitigation of the ray effects without a computing cost penalty

Numerical simulations of thermal conductivity in dissipative two-dimensional Yukawa systems.

Science.gov (United States)

Khrustalyov, Yu V; Vaulina, O S

2012-04-01

Numerical data on the heat transfer constants in two-dimensional Yukawa systems were obtained. Numerical study of the thermal conductivity and diffusivity was carried out for the equilibrium systems with parameters close to conditions of laboratory experiments with dusty plasma. For calculations of heat transfer constants the Green-Kubo formulas were used. The influence of dissipation (friction) on the heat transfer processes in nonideal systems was investigated. The approximation of the coefficient of thermal conductivity is proposed. Comparison of the obtained results to the existing experimental and numerical data is discussed.

A solution of the thermal neutron diffusion equation for a two-region cyclindrical system program for ODRA-1305 computer

International Nuclear Information System (INIS)

Drozdowicz, K.; Woznicka, U.

1982-01-01

The program in FORTRAN for the ODRA-1305 computer is described. The dependence of the decay constant of the thermal neutron flux upon the dimensions of the two-region concentric cylindrical system is the result of the program. The solution (with a constant neutron flux in the inner medium assumed) is generally obtained in the one-group diffusion approximation by the method of the perturbation calculation. However, the energy distribution of the thermal neutron flux and the diffusion cooling are taken into account. The program is written for the case when the outer medium is hydrogenous. The listing of the program and an example of calculation results are included. (author)

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments.

Science.gov (United States)

Miller, Marcelo E; Sztejnberg, Manuel L; González, Sara J; Thorp, Silvia I; Longhino, Juan M; Estryk, Guillermo

2011-12-01

A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comisión Nacional de Energía Atómica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Local mixed-field thermal neutron sensitivities and global thermal and mixed

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

Energy Technology Data Exchange (ETDEWEB)

Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo [Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429, Argentina and CONICET, Av. Rivadavia 1917, Ciudad de Buenos Aires 1033 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina)

2011-12-15

Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

International Nuclear Information System (INIS)

Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo

2011-01-01

Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and global

Calculation of the thermal neutron flux depression in the loop VISA-1

International Nuclear Information System (INIS)

Martinc, R.

1961-01-01

Among other applications, the VISA-1 loop is to be used for thermal load testing of materials. For this type of testing one should know the maximum power generated in the loop. This power is determined from the maximum thermal neutron flux in the VK-5 channel and mean flux depression in the fissile component of the loop. Thermal neutron flux depression is caused by neutron absorption in the components of the loop, shape of the components and neutron leaking through gaps as well as properties of the surrounding medium of the core. All these parameters were taken into account for calculating the depression of thermal neutron flux in the VISA-1 loop. Two group diffusion theory was used. Fast neutron from the fission in the loop and slowed down were taken into account. Depression of the thermal neutron flux is expressed by depression factor which represents the ratio of the mean thermal neutron flux in the fissile loop component and the thermal neutron flux in the VK-5 without the loop. Calculation error was estimated and it was recommended to determine the depression factor experimentally as well [sr

Thermal neutron imaging with rare-earth-ion-doped LiCaAlF{sub 6} scintillators and a sealed {sup 252}Cf source

Energy Technology Data Exchange (ETDEWEB)

Kawaguchi, Noriaki, E-mail: famicom@mail.tagen.tohoku.ac.jp [Tokuyama Corporation, Shibuya 3-chome, Shibuya-ku, Tokyo 150-8383 (Japan); IMRAM, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Yanagida, Takayuki; Fujimoto, Yutaka; Yokota, Yuui; Kamada, Kei [IMRAM, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Fukuda, Kentaro; Suyama, Toshihisa [Tokuyama Corporation, Shibuya 3-chome, Shibuya-ku, Tokyo 150-8383 (Japan); Watanabe, Kenichi; Yamazaki, Atsushi [Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Chani, Valery [IMRAM, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Yoshikawa, Akira [IMRAM, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); New Industry Creation Hatchery Center, Tohoku University, Sendai 980-8579 (Japan)

2011-10-01

Thermal neutron imaging with Ce-doped LiCaAlF{sub 6} crystals has been performed. The prototype of the neutron imager using a Ce-doped LiCaAlF{sub 6} scintillating crystal and a position sensitive photomultiplier tube (PSPMT) which had 64 multi-channel anode was developed. The Ce-doped LiCaAlF{sub 6} single crystal was grown by the Czochralski method. A plate with dimensions of a diameter of 50x2 mm{sup 2} was cut from the grown crystal, polished, and optically coupled to PSPMT by silicone grease. The {sup 252}Cf source (<1 MBq) was sealed with 43 mm of polyethylene for neutron thermalization. Alphabet-shaped Cd pieces with a thickness of 2 mm were used as a mask for the thermal neutrons. After corrections for the pedestals and gain of each pixel, we successfully obtained two-dimensional neutron images using Ce-doped LiCaAlF{sub 6}.

Longitudinal On-Column Thermal Modulation for Comprehensive Two-Dimensional Liquid Chromatography.

Science.gov (United States)

Creese, Mari E; Creese, Mathew J; Foley, Joe P; Cortes, Hernan J; Hilder, Emily F; Shellie, Robert A; Breadmore, Michael C

2017-01-17

Longitudinal on-column thermal modulation for comprehensive two-dimensional liquid chromatography is introduced. Modulation optimization involved a systematic investigation of heat transfer, analyte retention, and migration velocity at a range of temperatures. Longitudinal on-column thermal modulation was realized using a set of alkylphenones and compared to a conventional valve-modulator employing sample loops. The thermal modulator showed a reduced modulation-induced pressure impact than valve modulation, resulting in reduced baseline perturbation by a factor of 6; yielding a 6-14-fold improvement in signal-to-noise. A red wine sample was analyzed to demonstrate the potential of the longitudinal on-column thermal modulator for separation of a complex sample. Discrete peaks in the second dimension using the thermal modulator were 30-55% narrower than with the valve modulator. The results shown herein demonstrate the benefits of an active focusing modulator, such as reduced detection limits and increased total peak capacity.

A 3-D Thermal Analysis of the HANARO Cold Neutron Moderator Cell

International Nuclear Information System (INIS)

Han, Gee Y.; Kim, Heo Nil

2007-01-01

Fundamental studies on a thermal analysis of a cryogenic system such as a cold neutron source (CNS) have increased significantly for a successful CNS design in cold neutron research during recent years. A three-dimensional (3-D) thermal analysis model for the HANARO CNS was developed and used to accurately predict a temperature distribution between the hydrogen inside and the entire inner and outer surfaces of a moderator cell, whose moderator and cell walls are heated differently, under a steady-state operating condition by using the HEATING 7 code. The objective of this study is primarily to predict a temperature distribution through a heat flow in a cold neutron moderator cell heated from a nuclear heating and cooled by a cryogenic coolant. This paper presents satisfactory results of a steady-state temperature distribution in a cryogenic moderator cell. They are used to support the thermal stress analysis of the moderator cell walls and to provide a safe operation for the HANARO CNS facility

The diffusional pulsed cooling of the thermal neutron flux in small two-region systems. Monte Carlo Simulation

International Nuclear Information System (INIS)

Wiacek, U.

2006-06-01

The thermal neutron transport in small unhomogeneous system and namely in two- layers where the first one -outer moderator is of hydride type (polyethylene or plexiglas) and the second one - inner is made with other materials is investigated. The diffusional cooling of neutrons has been calculated by means of monte Carlo simulations using MCPN code. Because of un consistency of calculated and measured data the MCPN code library has been modified for polyethylene and plexiglas

Status of thermal neutron scattering data for graphite

International Nuclear Information System (INIS)

Mattes, M.; Keinert, J.

2005-07-01

At thermal neutron energies, the binding of the scattering nucleus in a solid, liquid, or gas affects the cross sections and the angular and energy distributions of the scattered neutrons. These effects are described in the thermal sub-library of evaluated files in File 7 of the ENDF-6 format. A re-evaluation of thermal neutron scattering data for carbon bound in graphite has been performed to investigate the impact of models (e.g., generalised frequency distributions) based on different experimental and theoretical data for the generation of scattering law data files S(α,β,T) and coherent elastic scattering data. Two phonon frequency distributions of graphite published in 2002 and 2004 were considered and the results compared with those based on the phonon spectra from Koppel et al. (published in 1968), on which the evaluations of ENDF/B-VI and JEFF-3.1 are based. The new frequency distributions were partly derived from ab initio simulations. Detailed comparisons with measurements of differential and integral neutron cross sections and other relevant data are reported. In addition, thermal MCNP data sets for use in the continuous Monte Carlo codes MCNP and MCNPX were generated from these evaluations for different temperatures. Calculated neutron spectra were found to be in good agreement with the measurements. (author)

Two-dimensional disruption thermal analysis code DREAM

International Nuclear Information System (INIS)

Yamazaki, Seiichiro; Kobayashi, Takeshi; Seki, Masahiro.

1988-08-01

When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.H.

1990-03-01

The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing, These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale of neutron shield from the cask. The test article was heated in an environmental prescribed by NRC regulations. Results of this second testing phase showed that all three materials are thermally acceptable

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.N.

1990-01-01

The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing. These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale section of neutron shield from the cask. The test article was heated in an environment prescribed by NRC regulations. Results of this second testing phase show that all three materials are thermally acceptable

Two-dimensional shielding benchmarks for iron at YAYOI, (1)

International Nuclear Information System (INIS)

Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.

The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)

Comparison of Thermal Neutron Flux Measured by Uranium 235 Fission Chamber and Rhodium Self-Powered Neutron Detector in MTR

International Nuclear Information System (INIS)

Fourmentel, D.; Filliatre, P.; Barbot, L.; Villard, J.-F.; Lyoussi, A.; Geslot, B.; Malo, J.-Y.; Carcreff, H.; Reynard-Carette, C.

2013-06-01

Thermal neutron flux is one of the most important nuclear parameter to be measured on-line in Material Testing Reactors (MTRs). In particular two types of sensors with different physical operating principles are commonly used: self-powered neutron detectors (SPND) and fission chambers with uranium 235 coating. This work aims to compare on one hand the thermal neutron flux evaluation given by these two types of sensors and on the other hand to compare these evaluations with activation dosimeter measurements, which are considered as the reference for absolute neutron flux assessment. This study was conducted in an irradiation experiment, called CARMEN-1, performed during 2012 in OSIRIS reactor (CEA Saclay - France). The CARMEN-1 experiment aims to improve the neutron and photon flux and nuclear heating measurements in MTRs. In this paper we focus on the thermal neutron flux measurements performed in CARMEN-1 experiment. The use of fission chambers to measure the absolute thermal neutron flux in MTRs is not very usual. An innovative calibration method for fission chambers operated in Campbell mode has been developed at the CEA Cadarache (France) and tested for the first time in the CARMEN-1 experiment. The results of these measurements are discussed, with the objective to measure with the best accuracy the thermal neutron flux in the future Jules Horowitz Reactor. (authors)

Numerical Simulations of Pillar Structured Solid State Thermal Neutron Detector Efficiency and Gamma Discrimination

Energy Technology Data Exchange (ETDEWEB)

Conway, A; Wang, T; Deo, N; Cheung, C; Nikolic, R

2008-06-24

This work reports numerical simulations of a novel three-dimensionally integrated, {sup 10}boron ({sup 10}B) and silicon p+, intrinsic, n+ (PIN) diode micropillar array for thermal neutron detection. The inter-digitated device structure has a high probability of interaction between the Si PIN pillars and the charged particles (alpha and {sup 7}Li) created from the neutron - {sup 10}B reaction. In this work, the effect of both the 3-D geometry (including pillar diameter, separation and height) and energy loss mechanisms are investigated via simulations to predict the neutron detection efficiency and gamma discrimination of this structure. The simulation results are demonstrated to compare well with the measurement results. This indicates that upon scaling the pillar height, a high efficiency thermal neutron detector is possible.

The use of diffusion theory to compute invasion effects for the pulsed neutron thermal decay time log

International Nuclear Information System (INIS)

Tittle, C.W.

1992-01-01

Diffusion theory has been successfully used to model the effect of fluid invasion into the formation for neutron porosity logs and for the gamma-gamma density log. The purpose of this paper is to present results of computations using a five-group time-dependent diffusion code on invasion effects for the pulsed neutron thermal decay time log. Previous invasion studies by the author involved the use of a three-dimensional three-group steady-state diffusion theory to model the dual-detector thermal neutron porosity log and the gamma-gamma density log. The five-group time-dependent code MGNDE (Multi-Group Neutron Diffusion Equation) used in this work was written by Ferguson. It has been successfully used to compute the intrinsic formation life-time correction for pulsed neutron thermal decay time logs. This application involves the effect of fluid invasion into the formation

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

International Nuclear Information System (INIS)

Liu, X.J.; Cheng, X.

2009-01-01

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

Semiconductor Thermal Neutron Detector

Directory of Open Access Journals (Sweden)

Toru Aoki

2014-02-01

Full Text Available The  CdTe  and  GaN  detector  with  a  Gd  converter  have  been developed  and  investigated  as  a  neutron  detector  for neutron  imaging.  The  fabricated  Gd/CdTe  detector  with  the  25  mm  thick  Gd  was  designed  on  the  basis  of  simulation results  of  thermal  neutron  detection  efficiency  and  spatial  resolution.  The  Gd/CdTe  detector  shows  the  detection  of neutron  capture  gamma  ray  emission  in  the  155Gd(n,  g156Gd,  157Gd(n,  g158Gd  and  113Cd(n,  g114Cd  reactions  and characteristic X-ray emissions due to conversion-electrons generated inside the Gd film. The observed efficient thermal neutron detection with the Gd/CdTe detector shows its promise in neutron radiography application. Moreover, a BGaN detector has also investigated to separate neutron signal from gamma-ray clearly. 

Dosimetry in Thermal Neutron Irradiation Facility at BMRR

Directory of Open Access Journals (Sweden)

Hu J.-P.

2016-01-01

Full Text Available Radiation dosimetry for Neutron Capture Therapy (NCT has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR. In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1 in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2 out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3 beam shutter upgrade to reduce strayed neutrons and gamma dose, (4 beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5 beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates to reduce prompt gamma and fast neutron doses, (6 sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7 holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4–7

Dosimetry in Thermal Neutron Irradiation Facility at BMRR

Energy Technology Data Exchange (ETDEWEB)

Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N.

2014-05-23

Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D₂O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7

A three-dimensional neutron transport benchmark solution

International Nuclear Information System (INIS)

Ganapol, B.D.; Kornreich, D.E.

1993-01-01

For one-group neutron transport theory in one dimension, several powerful analytical techniques have been developed to solve the neutron transport equation, including Caseology, Wiener-Hopf factorization, and Fourier and Laplace transform methods. In addition, after a Fourier transform in the transverse plane and formulation of a pseudo problem, two-dimensional (2-D) and three-dimensional (3-D) problems can be solved using the techniques specifically developed for the one-dimensional (1-D) case. Numerical evaluation of the resulting expressions requiring an inversion in the transverse plane have been successful for 2-D problems but becomes exceedingly difficult in the 3-D case. In this paper, we show that by using the symmetry along the beam direction, a 2-D problem can be transformed into a 3-D problem in an infinite medium. The numerical solution to the 3-D problem is then demonstrated. Thus, a true 3-D transport benchmark solution can be obtained from a well-established numerical solution to a 2-D problem

Two-dimensional Thermal Modeling of Lithium-ion Battery Cell Based on Electrothermal Impedance Spectroscopy

DEFF Research Database (Denmark)

Swierczynski, Maciej Jozef; Stroe, Daniel Loan; Knap, Vaclav

2016-01-01

Thermal modeling of lithium-ion batteries is gaining its importance together with increasing power density and compact design of the modern battery systems in order to assure battery safety and long lifetime. Thermal models of lithium-ion batteries are usually either expensive to develop...... and accurate or equivalent thermal circuit based with moderate accuracy and without spatial temperature distribution. This work presents initial results that can be used as a fundament for the cost-efficient development of the two-dimensional thermal model of lithium-ion battery based on multipoint...

Development of a large area, curved two-dimensional detector for single-crystal neutron diffraction studies

International Nuclear Information System (INIS)

Moon, Myung-Kook; Lee, Chang-Hee; Kim, Shin-Ae; Noda, Yukio

2013-01-01

A new type of two-dimensional curved position-sensitive neutron detector has been developed for a high-throughput single-crystal neutron diffractometer, which was designed to cover 110° horizontally and 56° vertically. The prototype curved detector covering 70° horizontally and 45° vertically was first developed to test the technical feasibility of the detector parameters, the internal anode and cathode structures for the curved shape, technical difficulties in the assembly procedure, and so on. Then, based on this experience, a full-scale curved detector with twice the active area of the prototype was fabricated with newly modified anode and cathode planes and optimized design parameters in terms of mechanical and electric properties. The detector was installed in a dedicated diffractometer at the ST3 beam port of the research reactor HANARO. In this paper, the fabrication and application of the prototype and a new larger-area curved position-sensitive neutron detector for single crystal diffraction is presented

Two-dimensional beam profiles and one-dimensional projections

Science.gov (United States)

Findlay, D. J. S.; Jones, B.; Adams, D. J.

2018-05-01

One-dimensional projections of improved two-dimensional representations of transverse profiles of particle beams are proposed for fitting to data from harp-type monitors measuring beam profiles on particle accelerators. Composite distributions, with tails smoothly matched on to a central (inverted) parabola, are shown to give noticeably better fits than single gaussian and single parabolic distributions to data from harp-type beam profile monitors all along the proton beam transport lines to the two target stations on the ISIS Spallation Neutron Source. Some implications for inferring beam current densities on the beam axis are noted.

Physics of epi-thermal boron neutron capture therapy (epi-thermal BNCT).

Science.gov (United States)

Seki, Ryoichi; Wakisaka, Yushi; Morimoto, Nami; Takashina, Masaaki; Koizumi, Masahiko; Toki, Hiroshi; Fukuda, Mitsuhiro

2017-12-01

The physics of epi-thermal neutrons in the human body is discussed in the effort to clarify the nature of the unique radiologic properties of boron neutron capture therapy (BNCT). This discussion leads to the computational method of Monte Carlo simulation in BNCT. The method is discussed through two examples based on model phantoms. The physics is kept at an introductory level in the discussion in this tutorial review.

Neutron Thermalization and Reactor Spectra. Vol. II. Proceedings of the Symposium on Neutron Thermalization and Reactor Spectra

International Nuclear Information System (INIS)

1968-01-01

Proceedings of a Symposium organized by the IAEA and held at Ann Arbor, Michigan, USA, 17 - 21 July 1967. The meeting was attended by 143 participants from 24 Member States and one international organization. Contents: (Vol.I) Theory of neutron thermalization (15 papers); Scattering law (20 papers); Angular, space, temperature and time dependence of neutron spectra (9 papers). (Vol.II) Measurement of thermal neutron spectra and spectral indices, and comparison with theory (17 papers); Time-dependent problems in neutron thermalization (12 papers). Each paper is in its original language (61 English, 1 French and 11 Russian) and is preceded by an abstract in English with one in the original language if this is not English. Discussions are in English.

Evaluation of thermal margin during BWR neutron flux oscillation

International Nuclear Information System (INIS)

Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

1992-01-01

Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

Analysis of cavity effect on space- and time-dependent fast and thermal neutron energy spectra

International Nuclear Information System (INIS)

Kudo, Katsuhisa; Narita, Masakuni; Ozawa, Yasutomo.

1975-01-01

The effects of the presence of a central cavity on the space- and time-dependent neutron energy spectra in both thermal and fast neutron systems are analyzed theoretically with use made of the multi-group one-dimensional time-dependent Ssub(n) method. The thermal neutron field is also analyzed for the case of a fundamental time eigenvalue problem with the time-dependent P 1 approximation. The cavity radius is variable, and the system radius for graphite is 120 cm and for the other materials 7 cm. From the analysis of the time-dependent Ssub(n) calculations in the non-multiplying systems of polythene, light water and graphite, cavity heating is the dominant effect for the slowing-down spectrum in the initial period following fast neutron burst, and when the slowing-down spectrum comes into the thermal energy region, cavity heating shifts to cavity cooling. In the multiplying system of 235 U, cavity cooling also takes place as the spectrum approaches equilibrium after the fast neutron burst is injected. The mechanism of cavity cooling is explained analytically for the case of thermal neutron field to illustrate its physical aspects, using the time-dependent P 1 approximation. An example is given for the case of light water. (auth.)

A preliminary inter-centre comparison study for photon, thermal neutron and epithermal neutron responses of two pairs of ionisation chambers used for BNCT

International Nuclear Information System (INIS)

Roca, Antoaneta; Liu, Yuan-Hao; Wojnecki, Cecile; Green, Stuart; Nievaart, Sander; Ghani, Zamir; Moss, Ray

2009-01-01

The dual ionisation chamber technique is the recommended method for mixed field dosimetry of epithermal neutron beams. This paper presents initial data from an ongoing inter-comparison study involving two identical pairs of ionisation chambers used at the BNCT facilities of Petten, NL and of University of Birmingham, UK. The goal of this study is to evaluate the photon, thermal neutron and epithermal neutron responses of both pairs of TE(TE) (Exradin T2 type) and Mg(Ar) (Exradin M2 type) ionisation chambers in similar experimental conditions. At this stage, the work has been completed for the M2 type chambers and is intended to be completed for the T2 type chambers in the near future.

Thermal Neutron Capture and Thermal Neutron Burn-up of K isomeric state of 177mLu: a way to the Neutron Super-Elastic Scattering cross section

International Nuclear Information System (INIS)

Roig, O.; Belier, G.; Meot, V.; Daugas, J.-M.; Romain, P.; Aupiais, J.; Jutier, Ch.; Le Petit, G.; Letourneau, A.; Marie, F.; Veyssiere, Ch.

2006-01-01

Thermal neutron radiative capture and burn-up measurements of the K isomeric state in 177Lu form part of an original method to indirectly obtain the neutron super-elastic scattering cross section at thermal energy. Neutron super-elastic scattering, also called neutron inelastic acceleration, occurs during the neutron collisions with an excited nuclear level. In this reaction, the nucleus could partly transfer its excitation energy to the scattered neutron

Flux distribution in phantom for biomedical use of beam-type thermal neutrons

International Nuclear Information System (INIS)

Aoki, Kazuhiko; Kobayashi, Tooru; Kanda, Keiji; Kimura, Itsuro

1985-01-01

For boron neutron capture therapy, the thermal neutron beam is worth using as therapeutic neutron irradiation without useless and unfavorable exposure of normal tissues around tumor and for microanalysis system to measure ppm-order 10 B concentrations in tissue and to search for the location of the metastasis of tumor. In the present study, the thermal neutron flux distribution in a phantom, when beam-type thermal neutrons were incident on it, was measured at the KUR Neutron Guide Tube. The measurements were carried out by two different methods using indium foil. The one is an ordinary foil activation technique by using the 115 In(n, γ) 116m 1 In reactions, while the other is to detect γ-rays from the 115 In(n, γ) 116m 2 In reactions during neutron irradiations with a handy-type Ge detector. The calculations with DOT 3.5 were performed to examine thermal neutron flux in the phantom for various beam size and phantom size. The experimental and calculated results are in good agreement and it is shown that the second type measurement has a potential for practical application as a new monitoring system of the thermal neutron flux in a living body for boron neutron capture therapy. (author)

Thermality and excited state Rényi entropy in two-dimensional CFT

Energy Technology Data Exchange (ETDEWEB)

Lin, Feng-Li [Department of Physics, National Taiwan Normal University,Taipei 11677, Taiwan (China); Wang, Huajia [Department of Physics, University of Illinois,Urbana-Champaign, IL 61801 (United States); Zhang, Jia-ju [Dipartimento di Fisica, Università degli Studi di Milano-Bicocca,Piazza della Scienza 3, I-20126 Milano (Italy); Theoretical Physics Division, Institute of High Energy Physics, Chinese Academy of Sciences,19B Yuquan Rd, Beijing 100049 (China); Theoretical Physics Center for Science Facilities, Chinese Academy of Sciences,19B Yuquan Rd, Beijing 100049 (China)

2016-11-21

We evaluate one-interval Rényi entropy and entanglement entropy for the excited states of two-dimensional conformal field theory (CFT) on a cylinder, and examine their differences from the ones for the thermal state. We assume the interval to be short so that we can use operator product expansion (OPE) of twist operators to calculate Rényi entropy in terms of sum of one-point functions of OPE blocks. We find that the entanglement entropy for highly excited state and thermal state behave the same way after appropriate identification of the conformal weight of the state with the temperature. However, there exists no such universal identification for the Rényi entropy in the short-interval expansion. Therefore, the highly excited state does not look thermal when comparing its Rényi entropy to the thermal state one. As the Rényi entropy captures the higher moments of the reduced density matrix but the entanglement entropy only the average, our results imply that the emergence of thermality depends on how refined we look into the entanglement structure of the underlying pure excited state.

Attenuation of thermal neutron through graphite

International Nuclear Information System (INIS)

Adib, M.; Ismaail, H.; Fathaallah, M.; Abbas, Y.; Habib, N.; Wahba, M.

2004-01-01

Calculation of the nuclear capture, thermal diffuse and Bragg scattering cross-sections as a function of graphite temperature and crystalline from for neutron energies from 1 me V< E<10 eV were carried out. Computer programs have been developed which allow calculation for the graphite hexagonal closed-pack structure in its polycrystalline form and pyrolytic one. I The calculated total cross-section for polycrystalline graphite were compared with the experimental values. An overall agreement is indicated between the calculated values and experimental ones. Agreement was also obtained for neutron cross-section measured for oriented pyrolytic graphite at room and liquid nitrogen temperatures. A feasibility study for use of graphite in powdered form as a cold neutron filter is details. The calculated attenuation of thermal neutrons through large mosaic pyrolytic graphite show that such crystals can be used effectively as second order filter of thermal neutron beams and that cooling improve their effectiveness

Two-dimensional geometrical corner singularities in neutron diffusion. Part 2: Application to the SNR-300 benchmark

International Nuclear Information System (INIS)

Cacuci, D.G.; Univ. of Karlsruhe; Kiefhaber, E.; Stehle, B.

1998-01-01

The explicit solution developed by Cacuci for the multigroup neutron diffusion equation at interior corners in two-dimensional two-region domains has been applied to the SNR-300 fast reactor prototype design to obtain the exact behavior of the multigroup fluxes at and around typical corners arising between absorber/fuel and follower/fuel assemblies. The calculations have been performed in hexagonal geometry using four energy groups, and the results clearly show that the multigroup fluxes are finite but not analytical at interior corners. In particular, already the first-order spatial derivatives of the multigroup fluxes become unbounded at the corners between follower and fuel assemblies. These results highlight the need to treat properly the influence of corners, both for the direct calculation and for the reconstruction of pointwise neutron flux and power distributions in heterogeneous reactor cores

Thermal neutron diffusion parameters in homogeneous mixtures

Energy Technology Data Exchange (ETDEWEB)

Drozdowicz, K.; Krynicka, E. [Institute of Nuclear Physics, Cracow (Poland)

1995-12-31

A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs.

Thermal neutron diffusion parameters in homogeneous mixtures

International Nuclear Information System (INIS)

Drozdowicz, K.; Krynicka, E.

1995-01-01

A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs

Thermal neutrons streaming in straight duct

International Nuclear Information System (INIS)

Jehouani, A.; Boulkheir, M.; Ichaoui, R.

2000-01-01

The neutron streaming in duct is due to two phenomena: a) direct propagation and b) reflection on duct wall. We have used the Monte Carlo method to evaluate the ratio of the reflected neutrons flux by the duct wall to the total flux at the exit of the duct for iron and aluminium. Ten neutrons energy groups are considered between 10 -5 eV and 10 eV. A Fortran program is developed to evaluate the neutron double differential albedo. It is shown that the two following approximations are largely justified: i) Three collisions in the duct wall are sufficient to attain the asymptotic limit of the multiscattered neutron double differential albedo ii) The points of entry and exit of the neutron in the duct wall may be considered the same for the multiscattered neutrons. For a punctual source at the mouth of the duct, we have determined the direct and the reflected part of the total thermal neutron flux at the exit of the duct for different lengths and different radius of the duct. For a punctual source, we have found that the major contribution to the total flux of neutrons at the exit is due to the neutron reflection by walls and the reflection contribution decreases when the neutron energy decreases. For a constant length of the duct, the reflected part decreases when the duct radius increases while for the disk shaped source we have found the opposite phenomena. The transmitted neutron flux distribution at the exit of the duct are determined for disk shaped source for different neutron energy and for different distance from the exit center. (author)

Three-Dimensional FIB/EBSD Characterization of Irradiated HfAl3-Al Composite

Energy Technology Data Exchange (ETDEWEB)

Hua, Zilong; Guillen, Donna Post; Harris, William; Ban, Heng

2016-09-01

A thermal neutron absorbing material, comprised of 28.4 vol% HfAl3 in an Al matrix, was developed to serve as a conductively cooled thermal neutron filter to enable fast flux materials and fuels testing in a pressurized water reactor. In order to observe the microstructural change of the HfAl3-Al composite due to neutron irradiation, an EBSD-FIB characterization approach is developed and presented in this paper. Using the focused ion beam (FIB), the sample was fabricated to 25µm × 25µm × 20 µm and mounted on the grid. A series of operations were carried out repetitively on the sample top surface to prepare it for scanning electron microscopy (SEM). First, a ~100-nm layer was removed by high voltage FIB milling. Then, several cleaning passes were performed on the newly exposed surface using low voltage FIB milling to improve the SEM image quality. Last, the surface was scanned by Electron Backscattering Diffraction (EBSD) to obtain the two-dimensional image. After 50 to 100 two-dimensional images were collected, the images were stacked to reconstruct a three-dimensional model using DREAM.3D software. Two such reconstructed three-dimensional models were obtained from samples of the original and post-irradiation HfAl3-Al composite respectively, from which the most significant microstructural change caused by neutron irradiation apparently is the size reduction of both HfAl3 and Al grains. The possible reason is the thermal expansion and related thermal strain from the thermal neutron absorption. This technique can be applied to three-dimensional microstructure characterization of irradiated materials.

Brazilian two-component TLD albedo neutron individual monitoring system

Energy Technology Data Exchange (ETDEWEB)

Martins, M.M., E-mail: marcelo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD), Av. Salvador Allende, s/n, CEP: 22780-160, Rio de Janeiro, RJ (Brazil); Mauricio, C.L.P., E-mail: claudia@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD), Av. Salvador Allende, s/n, CEP: 22780-160, Rio de Janeiro, RJ (Brazil); Fonseca, E.S. da, E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD), Av. Salvador Allende, s/n, CEP: 22780-160, Rio de Janeiro, RJ (Brazil); Silva, A.X. da, E-mail: ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao em Engenharia, COPPE/PEN Caixa Postal 68509, CEP: 21941-972, Rio de Janeiro, RJ (Brazil)

2010-12-15

Since 1983, Instituto de Radioprotecao e Dosimetria, Brazil, uses a TLD one-component albedo neutron monitor, which has a single different calibration factor specifically for each installation type. In order to improve its energy response, a two-component albedo monitor was developed, which measure the thermal neutron component besides the albedo one. The two-component monitor has been calibrated in reference neutron fields: thermal, five accelerator-produced monoenergetic beams (70, 144, 565, 1200 and 5000 keV) and five radionuclide sources ({sup 252}Cf, {sup 252}Cf(D{sub 2}O), {sup 241}Am-Be, {sup 241}Am-B and {sup 238}Pu-Be) at several distances. Since January 2008, mainly Brazilian workers who handle neutron sources at different distances and moderation, such as in well logging and calibration facilities are using it routinely.

Determination of cadmium in zinc ores by thermal neutron absorption analysis

International Nuclear Information System (INIS)

De Norre, L.; Op de Beeck, J.; Hoste, J.

1983-01-01

A method has been developed for routine determination of cadmium in zinc ores by thermal neutron absorption analysis, based on the attenuation of a thermal neutron flux passing through a neutron absorbing material. The thermal neutron flux in related to the 52 V activity induced in a vanadium detector, surrounded by pellets pressed from a mixture of powdered material with graphite. Besides cadmium, also the major constituents zinc, iron and sulfur contribute significantly to the total attenuation of the thermal neutron flux. Calibration lines for these elements are worked out. All irradiations are carried out for 200 s in the partially thermalized neutron flux of a 5 Ci 227 Ac-Be isotope neutron source. After a decay of 30 s, the 52 V activity of the vanadium detector is measured for 400 s with a NaI(Tl) scintillation detector. The analysis sequence, including the computation of the results from the counting data, is automated by means of a LSI-11 Microprocessor with 12Kx16 bit memory. Zinc ores, containing 0.02 to 1.45% cadmium, have been analyzed with a precision ranging from 12.6% to 0.54%, resp. As a test for the reliability of the method, two NBS standard reference materials were analyzed in the same way as the zinc ore samples. (author)

Three Dimensional Polarimetric Neutron Tomography of Magnetic Fields

DEFF Research Database (Denmark)

Sales, Morten; Strobl, Markus; Shinohara, Takenao

2018-01-01

Through the use of Time-of-Flight Three Dimensional Polarimetric Neutron Tomography (ToF 3DPNT) we have for the first time successfully demonstrated a technique capable of measuring and reconstructing three dimensional magnetic field strengths and directions unobtrusively and non-destructively wi......Through the use of Time-of-Flight Three Dimensional Polarimetric Neutron Tomography (ToF 3DPNT) we have for the first time successfully demonstrated a technique capable of measuring and reconstructing three dimensional magnetic field strengths and directions unobtrusively and non...... and reconstructed, thereby providing the proof-of-principle of a technique able to reveal hitherto unobtainable information on the magnetic fields in the bulk of materials and devices, due to a high degree of penetration into many materials, including metals, and the sensitivity of neutron polarisation to magnetic...... fields. The technique puts the potential of the ToF time structure of pulsed neutron sources to full use in order to optimise the recorded information quality and reduce measurement time....

Trial production of hyper-thermal neutron generator for Neutron Capture Therapy (NCT) and its radiation properties

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Toru

1999-01-01

In NCT, it was at first important to give a cancer portion to radiation dose required for its recovery. By finding out that whole cross-section of water comprising of a living body decreased monotonously with increase of neutron energy from about 100 barn against thermal neutron, became about 40 barn at about 0.5 eV and kept constant to 40 barn till at about 100 eV, application of thermal neutron shifted to higher temperature side, called Hyper thermal neutron, to NCT is proposed. The Hyper thermal neutron radiation can be expected to have similar controllability to that of the thermal neutron radiation. In 1977 fiscal year, a trial Hyper thermal neutron generator was produced on a base of up-to-date investigation results. As a part of property evaluation of the generator, evaluation of energy spectra in the Hyper thermal neutron generated at LINAC by TOF was conducted to confirm shift of the spectra to high temperature side. And, a Fantom experiment at KUR heavy water neutron radiation facility was also conducted to confirm effect of improvement in deep portion dose distribution. (G.K.)

Thermal neutron imaging through XRQA2 GAFCHROMIC films coupled with a cadmium radiator

Energy Technology Data Exchange (ETDEWEB)

Sacco, D. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); INAIL – DIT, Via di Fontana Candida n.1, 00040 Monteporzio Catone (Italy); Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Bortot, D. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Palomba, M. [ENEA Casaccia, Via Anguillarese, 301, S. Maria di Galeria, 00123 Roma (Italy); Pola, A. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); Gentile, A. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Strigari, L. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Pressello, C. [Department of Medical Physics, Azienda Ospedaliera San Camillo Forlanini, Circonvallazione Gianicolense 87, 00152 Roma (Italy); Soriani, A. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Gómez-Ros, J.M. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

2015-10-21

A simple and inexpensive method to perform passive thermal neutron imaging on large areas was developed on the basis of XRQA2 GAFCHROMIC films, commonly employed for quality assurance in radiology. To enhance their thermal neutron response, the sensitive face of film was coupled with a 1 mm thick cadmium radiator, forming a sandwich. By exchanging the order of Cd filter and sensitive film with respect to the incident neutron beam direction, two different configurations (beam-Cd-film and beam-film-Cd) were identified. These configurations were tested at thermal neutrons fluence values in the range 10{sup 9}–10{sup 10} cm{sup −2}, using the ex-core radial thermal neutron column of the ENEA Casaccia – TRIGA reactor. The results are presented in this work.

An analytical approach for a nodal formulation of a two-dimensional fixed-source neutron transport problem in heterogeneous medium

Energy Technology Data Exchange (ETDEWEB)

Basso Barichello, Liliane; Dias da Cunha, Rudnei [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Inst. de Matematica; Becker Picoloto, Camila [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Tres, Anderson [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Matematica Aplicada

2015-05-15

A nodal formulation of a fixed-source two-dimensional neutron transport problem, in Cartesian geometry, defined in a heterogeneous medium, is solved by an analytical approach. Explicit expressions, in terms of the spatial variables, are derived for averaged fluxes in each region in which the domain is subdivided. The procedure is an extension of an analytical discrete ordinates method, the ADO method, for the solution of the two-dimensional homogeneous medium case. The scheme is developed from the discrete ordinates version of the two-dimensional transport equation along with the level symmetric quadrature scheme. As usual for nodal schemes, relations between the averaged fluxes and the unknown angular fluxes at the contours are introduced as auxiliary equations. Numerical results are in agreement with results available in the literature.

Thermal neutron converter for irradiations with fission neutrons

International Nuclear Information System (INIS)

Wagner, F.M.; Kampfer, S.; Kastenmuller, A.; Waschkowski, W.; Bucherl, Th.; Kampfer, S.

2007-01-01

The new research reactor FRM II at Garching started operation in March 2004. The compact core is cooled by light water, and moderated by heavy water. Two fuel plates mounted in the heavy water tank convert thermal to fast neutrons. The fast neutron flux in the connected beam tube is up to 7 centre dot 10 8 s -1 cm -2 (depending on filters and collimation); the mean neutron energy is about 1.6 MeV. There are two irradiation rooms along the beam. The first is mainly used for medical therapy (MEDAPP facility), the second for materials characterization (NECTAR facility). At the former therapy facility RENT at the old research reactor FRM, the same beam quality was available until July 2000. Therefore, only a small program is run for the determination of the biological effectiveness of the new beam. The neutron and gamma dose rates in the medical beam are 0.54 and 0.20 Gy/min, respectively. The therapy facility MEDAPP is still under examination according to European regulations for medical devices. Full medical operation will start in 2007. The radiography and tomography facility NECTAR is in operation and aims at non-destructive inspection of objects up to 400 kg mass and 80 centre dot 80 centre dot 80 cm 3 in size. As for fission neutrons the macroscopic cross section of hydrogen is much higher than for other materials (e. g. Fe and Pb), one special application is the detection of hydrogen-containing materials (e. g. oil) in dense materials

Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

International Nuclear Information System (INIS)

Ganapol, B.D.; Kornreich, D.E.

1997-01-01

Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green's function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade

Three frequency modulated combination thermal neutron lifetime log and porosity

International Nuclear Information System (INIS)

Paap, H.J.; Arnold, D.M.; Smith, M.P.

1976-01-01

Methods are disclosed for measuring simultaneously the thermal neutron lifetime of the borehole fluid and earth formations in the vicinity of a well borehole, together with the formation porosity. A harmonically intensity modulated source of fast neutrons is used to irradiate the earth formations with fast neutrons at three different modulation frequencies. Intensity modulated clouds of thermal neutrons at each of the three modulation frequencies are detected by dual spaced detectors and the relative phase shift of the thermal neutrons with respect to the fast neutrons is determined at each of the three modulation frequencies at each detector. These measurements are then combined to determine simultaneously the thermal neutron decay time of the borehole fluid, the thermal neutron decay time of surrounding earth formation media and the porosity of the formation media

Thermal neutron calibration channel at LNMRI/IRD

International Nuclear Information System (INIS)

Astuto, A.; Salgado, A.P.; Lopes, R.T.; Leite, S.P.; Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W.

2014-01-01

The Brazilian Metrology Laboratory of Ionizing Radiations (LNMRI) standard thermal neutron flux facility was designed to provide uniform neutron fluence for calibration of small neutron detectors and individual dosemeters. This fluence is obtained by neutron moderation from four 241 Am-Be sources, each with 596 GBq, in a facility built with blocks of graphite/paraffin compound and high-purity carbon graphite. This study was carried out in two steps. In the first step, simulations using the MCNPX code on different geometric arrangements of moderator materials and neutron sources were performed. The quality of the resulting neutron fluence in terms of spectrum, cadmium ratio and gamma-neutron ratio was evaluated. In the second step, the system was assembled based on the results obtained on the simulations, and new measurements are being made. These measurements will validate the system, and other intercomparisons will ensure traceability to the International System of Units. The pile construction form using blocks allows distinct arrangements for new studies and possibilities of other LNMRI reference fields. The results can be predicted by the simulation used in this work. Different number of each type of blocks and sources can be used. The main difference observed between the final measurement and simulation results might be due to the difference in composition of paraffin blocks used in assembling the pile. In order to confirm the thermal neutron field and fluence rate in the central chamber (inside the channel) that will be used to irradiate small neutron detectors, it is necessary to use another quantification method such as the gold foils activation with measurement traceability. It will be performed in a future stage. (authors)

Method for determining thermal neutron decay times of earth formations

International Nuclear Information System (INIS)

Arnold, D.M.

1976-01-01

A method is disclosed for measuring the thermal neutron decay time of earth formations in the vicinity of a well borehole. A harmonically intensity modulated source of fast neutrons is used to irradiate the earth formations with fast neutrons at three different intensity modulation frequencies. The tangents of the relative phase angles of the fast neutrons and the resulting thermal neutrons at each of the three frequencies of modulation are measured. First and second approximations to the earth formation thermal neutron decay time are derived from the three tangent measurements. These approximations are then combined to derive a value for the true earth formation thermal neutron decay time

Activation measurements for thermal neutrons. Part J. Evaluation of thermal neutron transmission factors

International Nuclear Information System (INIS)

Egbert, Stephen D.

2005-01-01

In order to relate thermal neutron activation measurements in samples to the calculated free-in-air thermal neutron activation levels given in Chapter 3, use is made of sample transmission factors. Transmission factors account for the modification of the fluence and activation at each sample's in situ location. For the purposes of this discussion, the transmission factor (TF) is defined as the ratio of the in situ sample activation divided by the free-in-air (FIA) activation at a height of 1 m above ground at the same ground range. The procedures for calculation of TF's and example results are presented in this section. (author)

Thermal neutron albedo measurements for multilithic reflectors

International Nuclear Information System (INIS)

Mehboob, Khurram; Ahmed, Raheel; Ali, Majid; Tabassam, Uzma

2013-01-01

Highlights: • Measurement of thermal neuron albedo for multilithic reflectors. • Modeling of experiments in MATLAB. • Comparison of numerical calculated and experimental values. • Study of thermal neutron albedo in different multilayered shielding. - Abstract: An experimental measurement of the thermal neutron (0.025 eV) albedo (αth) has been carried out for multilithic shielding by using Am–Be neutron source and BF 3 detector. The measured saturation value for the thermal albedo of paraffin wax has been found to be 0.734 ± 0.020, which is in close agreement to the corresponding value 0.83 quoted in the literature. The thermal neutron albedo has been measured for the multilayered shielding in copper–wood, copper–aluminum, wood–paraffin and paraffin–iron combinations in horizontal geometric configurations. Modeling and numerical simulation have been carried out by developing a MATLAB code which solves the diffusion equation in order to calculate the experimental results. Good agreement has been found between the numerical calculated and experimental results. The uncertainties in the measurements have also been calculated based on error propagation of the underlying Poisson distribution

Biomedical applications of two- and three-dimensional deterministic radiation transport methods

International Nuclear Information System (INIS)

Nigg, D.W.

1992-01-01

Multidimensional deterministic radiation transport methods are routinely used in support of the Boron Neutron Capture Therapy (BNCT) Program at the Idaho National Engineering Laboratory (INEL). Typical applications of two-dimensional discrete-ordinates methods include neutron filter design, as well as phantom dosimetry. The epithermal-neutron filter for BNCT that is currently available at the Brookhaven Medical Research Reactor (BMRR) was designed using such methods. Good agreement between calculated and measured neutron fluxes was observed for this filter. Three-dimensional discrete-ordinates calculations are used routinely for dose-distribution calculations in three-dimensional phantoms placed in the BMRR beam, as well as for treatment planning verification for live canine subjects. Again, good agreement between calculated and measured neutron fluxes and dose levels is obtained

GEM-based thermal neutron beam monitors for spallation sources

International Nuclear Information System (INIS)

Croci, G.; Claps, G.; Caniello, R.; Cazzaniga, C.; Grosso, G.; Murtas, F.; Tardocchi, M.; Vassallo, E.; Gorini, G.; Horstmann, C.; Kampmann, R.; Nowak, G.; Stoermer, M.

2013-01-01

The development of new large area and high flux thermal neutron detectors for future neutron spallation sources, like the European Spallation Source (ESS) is motivated by the problem of 3 He shortage. In the framework of the development of ESS, GEM (Gas Electron Multiplier) is one of the detector technologies that are being explored as thermal neutron sensors. A first prototype of GEM-based thermal neutron beam monitor (bGEM) has been built during 2012. The bGEM is a triple GEM gaseous detector equipped with an aluminum cathode coated by 1μm thick B 4 C layer used to convert thermal neutrons to charged particles through the 10 B(n, 7 Li)α nuclear reaction. This paper describes the results obtained by testing a bGEM detector at the ISIS spallation source on the VESUVIO beamline. Beam profiles (FWHM x =31 mm and FWHM y =36 mm), bGEM thermal neutron counting efficiency (≈1%), detector stability (3.45%) and the time-of-flight spectrum of the beam were successfully measured. This prototype represents the first step towards the development of thermal neutrons detectors with efficiency larger than 50% as alternatives to 3 He-based gaseous detectors

Standard Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 The purpose of this test method is to define a general procedure for determining an unknown thermal-neutron fluence rate by neutron activation techniques. It is not practicable to describe completely a technique applicable to the large number of experimental situations that require the measurement of a thermal-neutron fluence rate. Therefore, this method is presented so that the user may adapt to his particular situation the fundamental procedures of the following techniques. 1.1.1 Radiometric counting technique using pure cobalt, pure gold, pure indium, cobalt-aluminum, alloy, gold-aluminum alloy, or indium-aluminum alloy. 1.1.2 Standard comparison technique using pure gold, or gold-aluminum alloy, and 1.1.3 Secondary standard comparison techniques using pure indium, indium-aluminum alloy, pure dysprosium, or dysprosium-aluminum alloy. 1.2 The techniques presented are limited to measurements at room temperatures. However, special problems when making thermal-neutron fluence rate measurements in high-...

Real-time thermal neutron radiographic detection systems

International Nuclear Information System (INIS)

Berger, H.; Bracher, D.A.

1976-01-01

Systems for real-time detection of thermal neutron images are reviewed. Characteristics of one system are presented; the data include contrast, resolution and speed of response over the thermal neutron intensity range 2.5 10 3 n/cm 2 -sec to 10 7 n/cm 2 -sec

Critical Behaviour of Pure and Site-Random Two Dimensional Antiferromagnets

DEFF Research Database (Denmark)

Birgenau, R. J.; Als-Nielsen, Jens Aage; Shirane, G.

1977-01-01

Quasielastic neutron scattering studies of the static critical behavior in the two-dimensional antiferromagnets K2NiF4, K2MnF4, and Rb2Mn0.5Ni0.5F4 are reported. For T......Quasielastic neutron scattering studies of the static critical behavior in the two-dimensional antiferromagnets K2NiF4, K2MnF4, and Rb2Mn0.5Ni0.5F4 are reported. For T...

Attenuation of Thermal Neutrons by Crystalline Silicon

International Nuclear Information System (INIS)

Adib, M.; Habib, N.; Ashry, A.; Fathalla, M.

2002-01-01

A simple formula is given which allows to calculate the contribution of the total neutron cross - section including the Bragg scattering from different (hkt) planes to the neutron * transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy .A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500μ eV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given

Parameters measurement for the thermal neutron beam in the thermal column hole of Xi’an pulse reactor

Institute of Scientific and Technical Information of China (English)

无

2010-01-01

The distribution of the neutron spectra in the thermal column hole of Xi’an pulse reactor was measured with the time-of-flight method.Compared with the thermal Maxwellian theory neutron spectra,the thermal neutron spectra measured is a little softer,and the average neutron energy of the experimental spectra is about 0.042±0.01 eV.The thermal neutron fluence rate at the front end of thermal column hole,measured with gold foil activation techniques,is about 1.18×105 cm-2 s-1.The standard uncertainty of the measured thermal neutron fluence is about 3%.The spectra-averaged cross section of 197Au(n,γ) determined by the experimental thermal neutron spectra is(92.8±0.93) ×10-24 cm2.

Microfabricated thermal modulator for comprehensive two-dimensional micro gas chromatography: design, thermal modeling, and preliminary testing.

Science.gov (United States)

Kim, Sung-Jin; Reidy, Shaelah M; Block, Bruce P; Wise, Kensall D; Zellers, Edward T; Kurabayashi, Katsuo

2010-07-07

In comprehensive two-dimensional gas chromatography (GC x GC), a modulator is placed at the juncture between two separation columns to focus and re-inject eluting mixture components, thereby enhancing the resolution and the selectivity of analytes. As part of an effort to develop a microGC x microGC prototype, in this report we present the design, fabrication, thermal operation, and initial testing of a two-stage microscale thermal modulator (microTM). The microTM contains two sequential serpentine Pyrex-on-Si microchannels (stages) that cryogenically trap analytes eluting from the first-dimension column and thermally inject them into the second-dimension column in a rapid, programmable manner. For each modulation cycle (typically 5 s for cooling with refrigeration work of 200 J and 100 ms for heating at 10 W), the microTM is kept approximately at -50 degrees C by a solid-state thermoelectric cooling unit placed within a few tens of micrometres of the device, and heated to 250 degrees C at 2800 degrees C s(-1) by integrated resistive microheaters and then cooled back to -50 degrees C at 250 degrees C s(-1). Thermal crosstalk between the two stages is less than 9%. A lumped heat transfer model is used to analyze the device design with respect to the rates of heating and cooling, power dissipation, and inter-stage thermal crosstalk as a function of Pyrex-membrane thickness, air-gap depth, and stage separation distance. Experimental results are in agreement with trends predicted by the model. Preliminary tests using a conventional capillary column interfaced to the microTM demonstrate the capability for enhanced sensitivity and resolution as well as the modulation of a mixture of alkanes.

Chemical warfare agents identification by thermal neutron detection

International Nuclear Information System (INIS)

Liu Boxue; Ai Xianyun; Tan Daoyuan; Zhang Dianqin

2000-01-01

The hydrogen concentration determination by thermal neutron detection is a non-destructive, fast and effective method to identify chemical warfare agents and TNT that contain different hydrogen fraction. When an isotropic neutron source is used to irradiate chemical ammunition, hydrogen atoms of the agent inside shell act as a moderator and slow down neutrons. The number of induced thermal neutrons depends mainly upon hydrogen content of the agent. Therefore measurement of thermal neutron influence can be used to determine hydrogen atom concentration, thereby to determine the chemical warfare agents. Under a certain geometry three calibration curves of count rate against hydrogen concentration were measured. According to the calibration curves, response of a chemical agent or TNT could be calculated. Differences of count rate among chemical agents and TNT for each kind of shells is greater than five times of standard deviations of count rate for any agent, so chemical agents or TNT could be identified correctly. Meanwhile, blast tube or liquid level of chemical warfare agent could affect the response of thermal neutron count rate, and thereby the result of identification. (author)

Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

International Nuclear Information System (INIS)

Vaiana, F.

2009-11-01

This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

A time-of-flight detector for thermal neutrons from radiotherapy Linacs

Energy Technology Data Exchange (ETDEWEB)

Conti, V. [Universita degli Studi di Milano and INFN di Milano (Italy)], E-mail: conti.Valentina@gmail.com; Bartesaghi, G. [Universita degli Studi di Milano and INFN di Milano (Italy); Bolognini, D.; Mascagna, V.; Perboni, C.; Prest, M.; Scazzi, S. [Universita dell' Insubria, Como and INFN di Milano (Italy); Mozzanica, A. [Universita degli Studi di Brescia and INFN sezione di Pavia (Italy); Cappelletti, P.; Frigerio, M.; Gelosa, S.; Monti, A.; Ostinelli, A. [Fisica Sanitaria, Ospedale S. Anna di Como (Italy); Giannini, G.; Vallazza, E. [INFN, sezione di Trieste and Universita degli Studi di Trieste (Italy)

2007-10-21

Boron Neutron Capture Therapy (BNCT) is a therapeutic technique exploiting the release of dose inside the tumour cell after a fission of a {sup 10}B nucleus following the capture of a thermal neutron. BNCT could be the treatment for extended tumors (liver, stomach, lung), radio-resistant ones (melanoma) or tumours surrounded by vital organs (brain). The application of BNCT requires a high thermal neutron flux (>5x10{sup 8}ncm{sup -2}s{sup -1}) with the correct energy spectrum (neutron energy <10keV), two requirements that for the moment are fulfilled only by nuclear reactors. The INFN PhoNeS (Photo Neutron Source) project is trying to produce such a neutron beam with standard radiotherapy Linacs, maximizing with a dedicated photo-neutron converter the neutrons produced by Giant Dipole Resonance by a high energy (>8MeV) photon beam. In this framework, we have developed a real-time detector to measure the thermal neutron time-of -flight to compute the flux and the energy spectrum. Given the pulsed nature of Linac beams, the detector is a single neutron counting system made of a scintillator detecting the photon emitted after the neutron capture by the hydrogen nuclei. The scintillator signal is sampled by a dedicated FPGA clock thus obtaining the exact arrival time of the neutron itself. The paper will present the detector and its electronics, the feasibility measurements with a Varian Clinac 1800/2100CD and comparison with a Monte Carlo simulation.

Thermal neutron pulsed parameters in non-hydrogenous systems. Experiment for lead grains

International Nuclear Information System (INIS)

Drozdowicz, K.; Gabanska, B.; Kosik, M.; Krynicka, E.; Woznicka, U.; Zaleski, T.

1997-01-01

In Czubek's method of measurement of the thermal neutron macroscopic absorption cross section a two-region geometry is applied where the investigated sample is surrounded by an external moderator. Both regions in the measurements made up till now were hydrogenous, which means the same type of the thermal neutron transport properties. In the paper a theoretical consideration to use non-hydrogenous materials as the samples is presented. Pulsed neutron measurements have been performed on homogeneous material in a geometry of the classic experiment with the variable geometric buckling. Two decay constants have been measured for different cylindrical samples of small lead grains (a lead shot). (author)

Thermal-hydraulic studies of the Advanced Neutron Source cold source

International Nuclear Information System (INIS)

Williams, P.T.; Lucas, A.T.

1995-08-01

The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design

Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

International Nuclear Information System (INIS)

D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

2003-01-01

A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

Energy Technology Data Exchange (ETDEWEB)

Slater, C.O.

1990-07-01

Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

Effect of thermal annealing on property changes of neutron-irradiated non-graphitized carbon materials and nuclear graphite

International Nuclear Information System (INIS)

Matsuo, Hideto

1991-06-01

Changes in dimension of non-graphitized carbon materials and nuclear graphite, and the bulk density, electrical resistivity, Young's modulus and thermal expansivity of nuclear graphite were studied after neutron irradiation at 1128-1483 K and the successive thermal annealing up to 2573 K. Carbon materials showed larger and anisotropic dimensional shrinkage than that of nuclear graphite after the irradiation. The irradiation-induced dimensional shrinkage of carbon materials decreased during annealing at temperatures from 1773 to 2023 K, followed by a slight increase at higher temperatures. On the other hand, the irradiated nuclear graphite hardly showed the changes in length, density and thermal expansivity under the thermal annealing, but the electrical resistivity and Young's modulus showed a gradual decrease with annealing temperature. It has been clarified that there exists significant difference in the effect of thermal annealing on irradiation-induced dimensional shrinkage between graphitized nuclear graphite and non-graphitized carbon materials. (author)

The CLYC-6 and CLYC-7 response to γ-rays, fast and thermal neutrons

International Nuclear Information System (INIS)

Giaz, A.; Pellegri, L.; Camera, F.; Blasi, N.; Brambilla, S.; Ceruti, S.; Million, B.; Riboldi, S.; Cazzaniga, C.; Gorini, G.; Nocente, M.; Pietropaolo, A.; Pillon, M.; Rebai, M.; Tardocchi, M.

2016-01-01

The crystal Cs 2 LiYCl 6 :Ce (CLYC) is a very interesting scintillator material because of its good energy resolution and its capability to identify γ-rays and fast/thermal neutrons. The crystal Cs 2 LiYCl 6 :Ce contains 6 Li and 35 Cl isotopes, therefore, it is possible to detect thermal neutrons through the reaction 6 Li(n, α)t while 35 Cl ions allow to measure fast neutrons through the reactions 35 Cl(n, p) 35 S and 35 Cl(n, α) 32 P. In this work two CLYC 1″×1″ crystals were used: the first crystal, enriched with 6 Li at 95% (CLYC-6) is ideal for thermal neutron measurements while the second one, enriched with 7 Li at >99% (CLYC-7) is suitable for fast neutron measurements. The response of CLYC scintillators was measured with different PMT models: timing or spectroscopic, with borosilicate glass or quartz window. The energy resolution, the neutron-γ discrimination and the internal activity are discussed. The capability of CLYC scintillators to discriminate γ rays from neutrons was tested with both thermal and fast neutrons. The thermal neutrons were measured with both detectors, using an AmBe source. The measurements of fast neutrons were performed at the Frascati Neutron Generator facility (Italy) where a deuterium beam was accelerated on a deuterium or on a tritium target, providing neutrons of 2.5 MeV or 14.1 MeV, respectively. The different sensitivity to thermal and fast neutrons of a CLYC-6 and of a CLYC-7 was additionally studied.

Comparison of preconditioned generalized conjugate gradient methods to two-dimensional neutron and photon transport equation

International Nuclear Information System (INIS)

Chen, G.S.; Yang, D.Y.

1998-01-01

We apply and compare the preconditioned generalized conjugate gradient methods to solve the linear system equation that arises in the two-dimensional neutron and photon transport equation in this paper. Several subroutines are developed on the basis of preconditioned generalized conjugate gradient methods for time-independent, two-dimensional neutron and photon transport equation in the transport theory. These generalized conjugate gradient methods are used: TFQMR (transpose free quasi-minimal residual algorithm) CGS (conjugate gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These subroutines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. The reasons to choose the generalized conjugate gradient methods are that the methods have better residual (equivalent to error) control procedures in the computation and have better convergent rate. The pointwise incomplete LU factorization ILU, modified pointwise incomplete LU factorization MILU, block incomplete factorization BILU and modified blockwise incomplete LU factorization MBILU are the preconditioning techniques used in the several testing problems. In Bi-CGSTAB, CGS, TFQMR and QMRCGSTAB method, we find that either CGS or Bi-CGSTAB method combined with preconditioner MBILU is the most efficient algorithm in these methods in the several testing problems. The numerical solution of flux by preconditioned CGS and Bi-CGSTAB methods has the same result as those from Cray computer, obtained by either the point successive relaxation method or the line successive relaxation method combined with Gaussian elimination

Study and development of new dosemeters for thermal neutrons

International Nuclear Information System (INIS)

Urena N, F.

1998-01-01

An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation (φ = 5 x 10 7 n/cm 2 -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence Φ = φ Δt ; where Δt = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

About possibilities of obtaining focused beams of thermal neutrons of radionuclide source

International Nuclear Information System (INIS)

Aripov, G.A.; Kurbanov, B.I.; Sulaymanov, N.T.; Ergashev, A.

2004-01-01

Full text: In the last years significant progress is achieved in development of neutron focusing methods (concentrating neutrons in a given direction and a small area). In this, main attention is given to focusing of neutron beams of reactor, particularly cold neutrons and their applications. [1,2]. However, isotope sources also let obtain intensive neutron beams and solve quite important (tasks) problems (e.g. neutron capture therapy for malignant tumors) [3], and an actual problems is focusing of neutrons. We developed a device on the basis of californium source of neutrons, allowing to obtain focused (preliminarily) beam of thermal neutrons with the aid of respective choice of moderators, reflectors and geometry of their disposition. Here, fast neutrons and gamma rays in the beam are minimized. With the aid of the model we developed on the basis of Monte-Carlo method, it is possible to modify aforementioned device and dynamics of output neutrons in wide energy range and analyze ways of optimization of neutron beams of isotope sources with different neutron outputs. Device of preliminary focusing of thermal neutrons can serve as a basis for further focus of neutrons using micro- and nano-capillar systems. It is known that, capillary systems performed with certain technology can form beam of thermal neutrons increasing its density by more than two orders of magnitude and effectively divert beams up to 20 o with length of system 15 cm

About possibilities of obtaining focused beams of thermal neutrons of radionuclide source

International Nuclear Information System (INIS)

Aripov, G.A.; Kurbanov, B.I.; Sulaymanov, N.T.; Ergashev, A.

2004-01-01

In the last years significant progress is achieved in development of neutron focusing methods (concentrating neutrons in a given direction and a small area). In this, main attention is given to focusing of neutron beams of reactor, particularly cold neutrons and their applications. [1,2]. However, isotope sources also let obtain intensive neutron beams and solve quite important (tasks) problems (e.g. neutron capture therapy for malignant tumors) [3], and an actual problems is focusing of neutrons. We developed a device on the basis of californium source of neutrons, allowing to obtain focused (preliminarily) beam of thermal neutrons with the aid of respective choice of moderators, reflectors and geometry of their disposition. Here, fast neutrons and gamma rays in the beam are minimized. With the aid of the model we developed on the basis of Monte-Carlo method, it is possible to modify aforementioned device and dynamics of output neutrons in wide energy range and analyze ways of optimization of neutron beams of isotope sources with different neutron outputs. Device of preliminary focusing of thermal neutrons can serve as a basis for further focus of neutrons using micro- and nano-capillary systems. It is known that, capillary systems performed with certain technology can form beam of thermal neutrons increasing its density by more than two orders of magnitude and effectively divert beams up to 20 o with length of system 15 cm. (author)

EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

Three-dimensional reconstruction of neutron, gamma-ray, and x-ray sources using spherical harmonic decomposition

Science.gov (United States)

Volegov, P. L.; Danly, C. R.; Fittinghoff, D.; Geppert-Kleinrath, V.; Grim, G.; Merrill, F. E.; Wilde, C. H.

2017-11-01

Neutron, gamma-ray, and x-ray imaging are important diagnostic tools at the National Ignition Facility (NIF) for measuring the two-dimensional (2D) size and shape of the neutron producing region, for probing the remaining ablator and measuring the extent of the DT plasmas during the stagnation phase of Inertial Confinement Fusion implosions. Due to the difficulty and expense of building these imagers, at most only a few two-dimensional projections images will be available to reconstruct the three-dimensional (3D) sources. In this paper, we present a technique that has been developed for the 3D reconstruction of neutron, gamma-ray, and x-ray sources from a minimal number of 2D projections using spherical harmonics decomposition. We present the detailed algorithms used for this characterization and the results of reconstructed sources from experimental neutron and x-ray data collected at OMEGA and NIF.

Assessment of fast and thermal neutron ambient dose equivalents around the KFUPM neutron source storage area using nuclear track detectors

Energy Technology Data Exchange (ETDEWEB)

Fazal-ur-Rehman [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)]. E-mail: fazalr@kfupm.edu.sa; Al-Jarallah, M.I. [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Abu-Jarad, F. [Radiation Protection Unit, Environmental Protection Department, Saudi Aramco, P. O. Box 13027, Dhahran 31311 (Saudi Arabia); Qureshi, M.A. [Center for Applied Physical Sciences, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)

2005-11-15

A set of five {sup 241}Am-Be neutron sources are utilized in research and teaching at King Fahd University of Petroleum and Minerals (KFUPM). Three of these sources have an activity of 16Ci each and the other two are of 5Ci each. A well-shielded storage area was designed for these sources. The aim of the study is to check the effectiveness of shielding of the KFUPM neutron source storage area. Poly allyl diglycol carbonate (PADC) Nuclear track detectors (NTDs) based fast and thermal neutron area passive dosimeters have been utilized side by side for 33 days to assess accumulated low ambient dose equivalents of fast and thermal neutrons at 30 different locations around the source storage area and adjacent rooms. Fast neutron measurements have been carried out using bare NTDs, which register fast neutrons through recoils of protons, in the detector material. NTDs were mounted with lithium tetra borate (Li{sub 2}B{sub 4}O{sub 7}) converters on their surfaces for thermal neutron detection via B10(n,{alpha})Li6 and Li6(n,{alpha})H3 nuclear reactions. The calibration factors of NTD both for fast and thermal neutron area passive dosimeters were determined using thermoluminescent dosimeters (TLD) with and without a polyethylene moderator. The calibration factors for fast and thermal neutron area passive dosimeters were found to be 1.33 proton tracks cm{sup -2}{mu}Sv{sup -1} and 31.5 alpha tracks cm{sup -2}{mu}Sv{sup -1}, respectively. The results show variations of accumulated dose with the locations around the storage area. The fast neutron dose equivalents rates varied from as low as 182nSvh{sup -1} up to 10.4{mu}Svh{sup -1} whereas those for thermal neutron ranged from as low as 7nSvh{sup -1} up to 9.3{mu}Svh{sup -1}. The study indicates that the area passive neutron dosimeter was able to detect dose rates as low as 7 and 182nSvh{sup -1} from accumulated dose for thermal and fast neutrons, respectively, which were not possible to detect with the available active neutron

Thermal neutron shield and method of manufacture

Science.gov (United States)

Brindza, Paul Daniel; Metzger, Bert Clayton

2013-05-28

A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

Contribution to the study of position sensitive detectors with high spatial resolution for thermal neutron detection

International Nuclear Information System (INIS)

Idrissi Fakhr-Eddine, Abdellah.

1978-01-01

With a view to improving the spatial resolution of the localization of thermal neutrons, the work covers four position sensitive detectors: - 800 cell multi-detectors (1 dimension), - linear 'Jeu de Jacquet' detectors (1 dimension) - Multi-detector XYP 128x128 (2 dimensions), - 'Jeu de Jacquet' detector with 2 dimensions. Mention is made of the various position finding methods known so far, as well as the reasons for selecting BF 3 as detector gas. A study is then made of the parameters of the multiwire chamber whose principle will form the basis of most of the position detecting appliances subsequently dealt with. Finally, a description is given of the detection tests of the thermal neutrons in the multiwire chamber depending on the pressure, a parameter that greatly affects the accuracy of the position finding. The single dimension position tests on two kinds of appliance, the 800 cell multi-detector for the wide angle diffraction studies, and the linear 'Jeu de Jacquet' detector designed for small angle diffraction are mentioned. A description is then given of two position appliances with two dimensions; the multi-detector XYP 128x128 and the two dimensional 'Jeu de Jacquet' detector. In the case of this latter detector, only the hoped for characteristics are indicated [fr

Thermal, epithermal and thermalized neutron attenuation properties of ilmenite-serpentine heat resistant concrete shield

International Nuclear Information System (INIS)

Kany, A.M.I.; El-Gohary, M.I.; Kamal, S.M.

1994-01-01

Experimental measurements were carried out to study the attenuation properties of low-energy neutrons transmitted through unheated and preheated barriers of heavy-weight, highly hydrated and heat-resistant concrete shields. The concrete shields under investigation have been prepared from naturally occurring ilmenite and serpentine Egyptian ores. A collimated beam obtained from an Am-Be source was used as a source of neutrons, while the measurements of total thermal, epithermal, and thermalized neutron fluxes were performed using a BF-3 detector, multichannel analyzer and Cd filter. Results show that the ilmenite-serpentine concrete proved to be a better thermal, epithermal and thermalized neutron attenuator than the ordinary concrete especially at a high temperature of concrete exposure. (Author)

High precision thermal neutron detectors

Energy Technology Data Exchange (ETDEWEB)

Radeka, V.; Schaknowski, N.A.; Smith, G.C.; Yu, B. [Brookhaven National Laboratory, Upton, NY (United States)

1994-12-31

Two-dimensional position sensitive detectors are indispensable in neutron diffraction experiments for determination of molecular and crystal structures in biology, solid-state physics and polymer chemistry. Some performance characteristics of these detectors are elementary and obvious, such as the position resolution, number of resolution elements, neutron detection efficiency, counting rate and sensitivity to gamma-ray background. High performance detectors are distinguished by more subtle characteristics such as the stability of the response (efficiency) versus position, stability of the recorded neutron positions, dynamic range, blooming or halo effects. While relatively few of them are needed around the world, these high performance devices are sophisticated and fairly complex, their development requires very specialized efforts. In this context, we describe here a program of detector development, based on {sup 3}He filled proportional chambers, which has been underway for some years at the Brookhaven National Laboratory. Fundamental approaches and practical considerations are outlined that have resulted in a series of high performance detectors with the best known position resolution, position stability, uniformity of response and reliability over time, for devices of this type.

Thermal neutron imaging in an active interrogation environment

International Nuclear Information System (INIS)

Vanier, P.E.; Forman, L.; Norman, D.R.

2009-01-01

We have developed a thermal-neutron coded-aperture imager that reveals the locations of hydrogenous materials from which thermal neutrons are being emitted. This imaging detector can be combined with an accelerator to form an active interrogation system in which fast neutrons are produced in a heavy metal target by means of excitation by high energy photons. The photo-induced neutrons can be either prompt or delayed, depending on whether neutronemitting fission products are generated. Provided that there are hydrogenous materials close to the target, some of the photo-induced neutrons slow down and emerge from the surface at thermal energies. These neutrons can be used to create images that show the location and shape of the thermalizing materials. Analysis of the temporal response of the neutron flux provides information about delayed neutrons from induced fission if there are fissionable materials in the target. The combination of imaging and time-of-flight discrimination helps to improve the signal-to-background ratio. It is also possible to interrogate the target with neutrons, for example using a D-T generator. In this case, an image can be obtained from hydrogenous material in a target without the presence of heavy metal. In addition, if fissionable material is present in the target, probing with fast neutrons can stimulate delayed neutrons from fission, and the imager can detect and locate the object of interest, using appropriate time gating. Operation of this sensitive detection equipment in the vicinity of an accelerator presents a number of challenges, because the accelerator emits electromagnetic interference as well as stray ionizing radiation, which can mask the signals of interest.

Boundary element methods applied to two-dimensional neutron diffusion problems

International Nuclear Information System (INIS)

Itagaki, Masafumi

1985-01-01

The Boundary element method (BEM) has been applied to two-dimensional neutron diffusion problems. The boundary integral equation and its discretized form have been derived. Some numerical techniques have been developed, which can be applied to critical and fixed-source problems including multi-region ones. Two types of test programs have been developed according to whether the 'zero-determinant search' or the 'source iteration' technique is adopted for criticality search. Both programs require only the fluxes and currents on boundaries as the unknown variables. The former allows a reduction in computing time and memory in comparison with the finite element method (FEM). The latter is not always efficient in terms of computing time due to the domain integral related to the inhomogeneous source term; however, this domain integral can be replaced by the equivalent boundary integral for a region with a non-multiplying medium or with a uniform source, resulting in a significant reduction in computing time. The BEM, as well as the FEM, is well suited for solving irregular geometrical problems for which the finite difference method (FDM) is unsuited. The BEM also solves problems with infinite domains, which cannot be solved by the ordinary FEM and FDM. Some simple test calculations are made to compare the BEM with the FEM and FDM, and discussions are made concerning the relative merits of the BEM and problems requiring future solution. (author)

Development of an Advanced Two-Dimensional Thermal Model for Large size Lithium-ion Pouch Cells

International Nuclear Information System (INIS)

Samba, Ahmadou; Omar, Noshin; Gualous, Hamid; Firouz, Youssef; Van den Bossche, Peter; Van Mierlo, Joeri; Boubekeur, Tala Ighil

2014-01-01

In this work, a LiFePO4/graphite lithium-ion pouch cell with a rated capacity of 45Ah has been used and a two dimensional thermal model is developed to predict the cell temperature distribution over the surface of the battery, this model requires less input parameters and still has high accuracy. The used input parameters are the heat generation and thermal properties. The ANSYS FLUENT software has been used to solve the models. In addition, a new estimation tool has been developed for estimation of the thermal model parameters. Furthermore, the thermal behavior of the proposed battery has been investigated at different environmental conditions as well as during the abuse conditions. Thermal runaway is investigated in depth by the model

Development of temperature related thermal neutron scattering database for MCNP

International Nuclear Information System (INIS)

Mei Longwei; Cai Xiangzhou; Jiang Dazhen; Chen Jingen; Guo Wei

2013-01-01

Based on ENDF/B-Ⅶ neutron library, the thermal neutron scattering library S(α, β) for molten salt reactor moderators was developed. The temperatures of this library were chose as the characteristic temperature of the molten salt reactor. The cross section of the thermal neutron scattering of ACE format was investigated, and this library was also validated by the benchmarks of ICSBEP. The uncertainties shown in the validation were in reasonable range when compared with the thermal neutron scattering library tmccs which included in the MCNP data library. It was proved that the thermal neutron scattering library processed in this study could be used in the molten salt reactor design. (authors)

Thermal neutron standard fields with the KUR heavy water facility

International Nuclear Information System (INIS)

Kanda, K.; Kobayashi, K.; Shibata, T.

1978-01-01

A heavy water facility attached to the KUR (Kyoto University Reactor, swimming pool type, 5 MW) yields pure thermal neutrons in the Maxwellian distribution. The facility is faced to the core of KUR and it contains about 2 tons of heavy water. The thickness of the layer is about 140 cm. The neutron spectrum was measured with the time of flight technique using a fast chopper. The measured spectrum was in good agreement with the Maxwellian distribution in all energy region for thermal neutrons. The neutron temperature was slightly higher than the heavy water temperature. The contamination of epithermal and fast neutrons caused by photo-neutrons of the γ-n reaction of heavy water was very small. The maximum intensity of thermal neutrons is 3x10 11 n/cm 2 sec. When the bismuth scatterer is attached, the gamma rays contamination is eliminated by the ratio of 0.05 of gamma rays to neutrons in rem. This standard neutron field has been used for such experiments as thermal neutron cross section measurement, detector calibration, activation analysis, biomedical purposes etc. (author)

Improved adaptive genetic algorithm with sparsity constraint applied to thermal neutron CT reconstruction of two-phase flow

Science.gov (United States)

Yan, Mingfei; Hu, Huasi; Otake, Yoshie; Taketani, Atsushi; Wakabayashi, Yasuo; Yanagimachi, Shinzo; Wang, Sheng; Pan, Ziheng; Hu, Guang

2018-05-01

Thermal neutron computer tomography (CT) is a useful tool for visualizing two-phase flow due to its high imaging contrast and strong penetrability of neutrons for tube walls constructed with metallic material. A novel approach for two-phase flow CT reconstruction based on an improved adaptive genetic algorithm with sparsity constraint (IAGA-SC) is proposed in this paper. In the algorithm, the neighborhood mutation operator is used to ensure the continuity of the reconstructed object. The adaptive crossover probability P c and mutation probability P m are improved to help the adaptive genetic algorithm (AGA) achieve the global optimum. The reconstructed results for projection data, obtained from Monte Carlo simulation, indicate that the comprehensive performance of the IAGA-SC algorithm exceeds the adaptive steepest descent-projection onto convex sets (ASD-POCS) algorithm in restoring typical and complex flow regimes. It especially shows great advantages in restoring the simply connected flow regimes and the shape of object. In addition, the CT experiment for two-phase flow phantoms was conducted on the accelerator-driven neutron source to verify the performance of the developed IAGA-SC algorithm.

Modelling of the thermal parameters of high-power linear laser-diode arrays. Two-dimensional transient model

International Nuclear Information System (INIS)

Bezotosnyi, V V; Kumykov, Kh Kh

1998-01-01

A two-dimensional transient thermal model of an injection laser is developed. This model makes it possible to analyse the temperature profiles in pulsed and cw stripe lasers with an arbitrary width of the stripe contact, and also in linear laser-diode arrays. This can be done for any durations and repetition rates of the pump pulses. The model can also be applied to two-dimensional laser-diode arrays operating quasicontinuously. An analysis is reported of the influence of various structural parameters of a diode array on the thermal regime of a single laser. The temperature distributions along the cavity axis are investigated for different variants of mounting a crystal on a heat sink. It is found that the temperature drop along the cavity length in cw and quasi-cw laser diodes may exceed 20%. (lasers)

COBRA/TRAC analysis of two-dimensional thermal-hydraulic behavior in SCTF reflood tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Ohnuki, Akira; Sobajima, Makoto; Adachi, Hiromichi

1987-01-01

The effects of radial power distribution and non-uniform upper plenum water accumulation on thermal-hydraulic behavior in the core were observed in the reflood tests with Slab Core Test Facility (SCTF). In order to examine the predictability of these two effects by a multi-dimensional analysis code, the COBRA/TRAC calculations were made. The calculated results indicated that the heat transfer enhancement in high power bundles above quench front was caused by high vapor flow rate in those bundles due to the radial power distribution. On the other hand, the heat transfer degradation in the peripheral bundles under the condition of non-uniform upper plenum water accumulation was caused by the lower flow rates of vapor and entrained liquid above the quench front in those bundles by the reason that vapor concentrated in the center bundles due to the cross flow induced by the horizontal pressure gradient in the core. The above-mentioned two-dimensional heat transfer behaviors calculated with the COBRA/TRAC code is similar to those observed in SCTF tests and therefore those calculations are useful to investigate the mechanism of the two-dimensional effects in SCTF reflood tests. (author)

Thermalization of monoenergetic neutrons in a concrete room

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado, G.A. [UAZ, A.P. 336, 98000 Zacatecas (Mexico); Iniguez, M.P.; Martin M, A. [Universidad de Valladolid, (Spain)

2006-07-01

The thermalization of neutrons from monoenergetic neutron sources in a concrete room has been studied. During calibration of neutron detectors it is mandatory to make corrections due to neutron scattering produced by the room walls, therefore this factor must be known in advance. The scattered neutrons are thermalized and produce a neutron field that is directly proportional to source strength and inversely proportional to room total wall-surfaces, the proportional coefficient has been calculated for neutrons whose energy goes from 1 eV to 20 MeV. This coefficient was calculated using Monte Carlo methods for 150, 200 and 300 cm-radius spherical cavity, where monoenergetic neutrons were located at the center, along the spherical cavity radius neutron spectra were calculated at several source-to-detector distances inside the cavity. The obtained coefficient is almost three times larger than the factor normally utilized. (Author)

Joint estimation of the fast and thermal components of a high neutron flux with a two on-line detector system

International Nuclear Information System (INIS)

Filliatre, P.; Oriol, L.; Jammes, C.; Vermeeren, L.

2009-01-01

A fission chamber with a 242 Pu deposit is the best suited detector for on-line measurements of the fast component of a high neutron flux (∼10 14 ncm -2 s -1 or more) with a significant thermal component. To get the fast flux, it is, however, necessary to subtract the contribution of the thermal neutrons, which increases with fluence because of the evolution of the isotopic content of the deposit. This paper presents an algorithm that permits, thanks to measurements provided by a 242 Pu fission chamber and a detector for thermal neutrons, to estimate the thermal and the fast flux at any time. An implementation allows to test it with simulated data.

Simultaneous thermal neutron decay time and porosity logging system

International Nuclear Information System (INIS)

Shultz, W.E.

1980-01-01

A method for simultaneously determining the porosity and thermal neutron capture cross-section of earth formations in the vicinity of a well borehole is claimed. It comprises the following steps: passing a well tool into a cased well borehole. The tool has a pulsed source of fast neutrons, a combination fast neutron and gamma ray detector and an epithermal neutron detector; repetitively irradiating the earth formations in the vicinity of the borehole with bursts of fast neutrons; detecting the fast neutron and epithermal neutron populations in the borehole (during the neutron bursts) and generating first and second measurement signals; detecting for second and third time intervals during the time between the neutron bursts, the gamma radiation present in the borehole due to the capture of thermalized neutrons by the nuclei of elements comprising the earth formations and generating third and fourth measurement signals; and combining the first and second measurement signals according to a predetermined relationship to derive an indication of the porosity of the earth formations and combining the third and fourth measurement signals to derive an indication of the thermal neutron capture cross-section of the earth formations

Limitations to the use of two-dimensional thermal modeling of a nuclear waste repository

International Nuclear Information System (INIS)

Davis, B.W.

1979-01-01

Thermal modeling of a nuclear waste repository is basic to most waste management predictive models. It is important that the modeling techniques accurately determine the time-dependent temperature distribution of the waste emplacement media. Recent modeling studies show that the time-dependent temperature distribution can be accurately modeled in the far-field using a 2-dimensional (2-D) planar numerical model; however, the near-field cannot be modeled accurately enough by either 2-D axisymmetric or 2-D planar numerical models for repositories in salt. The accuracy limits of 2-D modeling were defined by comparing results from 3-dimensional (3-D) TRUMP modeling with results from both 2-D axisymmetric and 2-D planar. Both TRUMP and ADINAT were employed as modeling tools. Two-dimensional results from the finite element code, ADINAT were compared with 2-D results from the finite difference code, TRUMP; they showed almost perfect correspondence in the far-field. This result adds substantially to confidence in future use of ADINAT and its companion stress code ADINA for thermal stress analysis. ADINAT was found to be somewhat sensitive to time step and mesh aspect ratio. 13 figures, 4 tables

One-dimensional two-phase thermal hydraulics (ENSTA course)

International Nuclear Information System (INIS)

Olive, J.

1995-11-01

This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends

Investigation of the dispersion of phonon modes in CdI2 single crystals by a method of inelastic scattering of thermal neutrons

International Nuclear Information System (INIS)

Piroga, S.A.

2001-01-01

Experimental observation using a method of inelastic scattering of thermal neutrons the longitudinal phonons in the G-Z, G-X and G-L directions in CdI 2 singe crystal has been obtained. The phonon subsystem observed in the case of CdI 2 single crystals is two dimensional. This is because of the fact that interlayer interactions are weak in compare to intra layer interactions

Thermal neutron detectors based on complex oxide crystals

CERN Document Server

Ryzhikov, V; Volkov, V; Chernikov, V; Zelenskaya, O

2002-01-01

The ways of improvement of spectrometric quality of CWO and GSO crystals have been investigated with the aim of their application in thermal neutron detectors based on radiation capture reactions. The efficiency of the neutron detection by these crystals was measured, and the obtained data were compared with the results for sup 6 LiI(Tl) crystals. It is shown that the use of complex oxide crystals and neutron-absorption filters for spectrometry of thermal and resonance neutrons could be a promising method in combination with computer data processing. Numerical calculations are reported for spectra of gamma-quanta due to radiation capture of the neutrons. To compensate for the gamma-background lines, we used a crystal pair of heavy complex oxides with different sensitivity to neutrons.

Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

Energy Technology Data Exchange (ETDEWEB)

Soltes, Jaroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague, (Czech Republic); Viererbl, Ladislav; Lahodova, Zdena; Koleska, Michal; Vins, Miroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic)

2015-07-01

In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which has a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values

Thermal conductivity in one-dimensional nonlinear systems

Science.gov (United States)

Politi, Antonio; Giardinà, Cristian; Livi, Roberto; Vassalli, Massimo

2000-03-01

Thermal conducitivity of one-dimensional nonlinear systems typically diverges in the thermodynamic limit, whenever the momentum is conserved (i.e. in the absence of interactions with an external substrate). Evidence comes from detailed studies of Fermi-Pasta-Ulam and diatomic Toda chains. Here, we discuss the first example of a one-dimensional system obeying Fourier law : a chain of coupled rotators. Numerical estimates of the thermal conductivity obtained by simulating a chain in contact with two thermal baths at different temperatures are found to be consistent with those ones based on linear response theory. The dynamics of the Fourier modes provides direct evidence of energy diffusion. The finiteness of the conductivity is traced back to the occurrence of phase-jumps. Our conclusions are confirmed by the analysis of two variants of the rotator model.

Using thermalizers in measuring 'Ukryttia' object's FCM neutron fluxes

CERN Document Server

Krasnyanskaya, O G; Odinokin, G I; Pavlovich, V N

2003-01-01

The results of research of a thermalizer (heater) width influence on neutron thermalization efficiency during FCM neutron flux measuring in the 'Ukryttia' are described. The calculations of neutron flux densities were performed by the Monte-Carlo method with the help of computer code MCNP-4C for FCM different models.Three possible installations of detectors were considered: on FCM surface,inside the FCM, and inside the concrete under the FCM layer. It was shown,that in order to increase the sensitivity of neutron detectors in intermediate and fast neutrons field,and consequently, to decrease the dependence of the readings of spectral distribution of neutron flux,it is necessary to position the detector inside the so-called thermalizer or heater. The most reasonable application of thick 'heaters' is the situation, when the detector is placed on FCM surface.

On some classes of two-dimensional local models in discrete two-dimensional monatomic FPU lattice with cubic and quartic potential

International Nuclear Information System (INIS)

Quan, Xu; Qiang, Tian

2009-01-01

This paper discusses the two-dimensional discrete monatomic Fermi–Pasta–Ulam lattice, by using the method of multiple-scale and the quasi-discreteness approach. By taking into account the interaction between the atoms in the lattice and their nearest neighbours, it obtains some classes of two-dimensional local models as follows: two-dimensional bright and dark discrete soliton trains, two-dimensional bright and dark line discrete breathers, and two-dimensional bright and dark discrete breather. (condensed matter: structure, thermal and mechanical properties)

LANSCE steady state unperturbed thermal neutron fluxes at 100 μA

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

The ''maximum'' unperturbed, steady state thermal neutron flux for LANSCE is calculated to be 2 /times/ 10 13 n/cm 2 -s for 100 μA of 800-MeV protons. This LANSCE neutron flux is a comparable entity to a steady state reactor thermal neutron flux. LANSCE perturbed steady state thermal neutron fluxes have also been calculated. Because LANSCE is a pulsed neutron source, much higher ''peak'' (in time) neutron fluxes can be generated than at a steady state reactor source. 5 refs., 5 figs

Joint estimation of the fast and thermal components of a high neutron flux with a two on-line detector system

Energy Technology Data Exchange (ETDEWEB)

Filliatre, P. [CEA, DEN, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Laboratoire Commun d' Instrumentation CEA-SCK-CEN (France)], E-mail: philippe.filliatre@cea.fr; Oriol, L.; Jammes, C. [CEA, DEN, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Laboratoire Commun d' Instrumentation CEA-SCK-CEN (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Laboratoire Commun d' Instrumentation CEA-SCK-CEN (France)

2009-05-21

A fission chamber with a {sup 242}Pu deposit is the best suited detector for on-line measurements of the fast component of a high neutron flux ({approx}10{sup 14}ncm{sup -2}s{sup -1} or more) with a significant thermal component. To get the fast flux, it is, however, necessary to subtract the contribution of the thermal neutrons, which increases with fluence because of the evolution of the isotopic content of the deposit. This paper presents an algorithm that permits, thanks to measurements provided by a {sup 242}Pu fission chamber and a detector for thermal neutrons, to estimate the thermal and the fast flux at any time. An implementation allows to test it with simulated data.

Response of six neutron survey meters in mixed fields of fast and thermal neutrons.

Science.gov (United States)

Kim, S I; Kim, B H; Chang, I; Lee, J I; Kim, J L; Pradhan, A S

2013-10-01

Calibration neutron fields have been developed at KAERI (Korea Atomic Energy Research Institute) to study the responses of commonly used neutron survey meters in the presence of fast neutrons of energy around 10 MeV. The neutron fields were produced by using neutrons from the (241)Am-Be sources held in a graphite pile and a DT neutron generator. The spectral details and the ambient dose equivalent rates of the calibration fields were established, and the responses of six neutron survey meters were evaluated. Four single-moderator-based survey meters exhibited an under-responses ranging from ∼9 to 55 %. DINEUTRUN, commonly used in fields around nuclear reactors, exhibited an over-response by a factor of three in the thermal neutron field and an under-response of ∼85 % in the mixed fields. REM-500 (tissue-equivalent proportional counter) exhibited a response close to 1.0 in the fast neutron fields and an under-response of ∼50 % in the thermal neutron field.

Test of a two-dimensional neutron spin analyzer

International Nuclear Information System (INIS)

Falus, Peter; Vorobiev, Alexei; Krist, Thomas

2006-01-01

The aim of this measurement was to test the new large-area spin polarization analyzer for the EVA-SERGIS beamline at Institute Laue Langevin (ILL). The spin analyzer, which was built in Berlin selects one of the two spin states of a neutron beam of wavelength 5.5 A impinging on a horizontal sample and reflected or scattered from the sample. The spin is analyzed for all neutrons scattered into a detector with an area of 190 mmx190 mm positioned 2.7 m behind the sample, thus covering an angular interval of 4 o x4 o . The tests were done at the HMI V14 beamline followed by tests at the EVA beamline at ILL. The transmission for the two spin components, the flipping ratio and small angle scattering were recorded while scanning the incoming beam on the analyzer. It was clearly visible, that due to the stacked construction the intensity is blocked at regular intervals. Careful inspection shows that the transmission of the good spin component is more than 0.72 for 60% of the detector area and the corrected flipping ratio is more than 47 for 60% of the detector area. Although some small-angle scattering is visible, it is notable that this analyzer design has small scattering intensities

Test of a two-dimensional neutron spin analyzer

Science.gov (United States)

Falus, Péter; Vorobiev, Alexei; Krist, Thomas

2006-11-01

The aim of this measurement was to test the new large-area spin polarization analyzer for the EVA-SERGIS beamline at Institute Laue Langevin (ILL). The spin analyzer, which was built in Berlin selects one of the two spin states of a neutron beam of wavelength 5.5 Å impinging on a horizontal sample and reflected or scattered from the sample. The spin is analyzed for all neutrons scattered into a detector with an area of 190 mm×190 mm positioned 2.7 m behind the sample, thus covering an angular interval of 4°×4°. The tests were done at the HMI V14 beamline followed by tests at the EVA beamline at ILL. The transmission for the two spin components, the flipping ratio and small angle scattering were recorded while scanning the incoming beam on the analyzer. It was clearly visible, that due to the stacked construction the intensity is blocked at regular intervals. Careful inspection shows that the transmission of the good spin component is more than 0.72 for 60% of the detector area and the corrected flipping ratio is more than 47 for 60% of the detector area. Although some small-angle scattering is visible, it is notable that this analyzer design has small scattering intensities.

Neutron--neutron logging

International Nuclear Information System (INIS)

Allen, L.S.

1977-01-01

A borehole logging tool includes a steady-state source of fast neutrons, two epithermal neutron detectors, and two thermal neutron detectors. A count rate meter is connected to each neutron detector. A first ratio detector provides an indication of the porosity of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two epithermal neutron detectors. A second ratio detector provides an indication of both porosity and macroscopic absorption cross section of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two thermal neutron detectors. By comparing the signals of the two ratio detectors, oil bearing zones and salt water bearing zones within the formation being logged can be distinguished and the amount of oil saturation can be determined. 6 claims, 2 figures

Application of the three-dimensional transport code to analysis of the neutron streaming experiment

International Nuclear Information System (INIS)

Chatani, K.; Slater, C.O.

1990-01-01

The neutron streaming through an experimental mock-up of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway was recalculated with a three-dimensional discrete ordinates code. The experiment was conducted at the Tower Shielding Facility at Oak Ridge National Laboratory in 1976 and 1977. The measurement of the neutron flux, using Bonner ball detectors, indicated nine orders of attenuation in the empty pipeway, which contained two 90-deg bends and was surrounded by concrete walls. The measurement data were originally analyzed using the DOT3.5 two-dimensional discrete ordinates radiation transport code. However, the results did not agree with measurement data at the bend because of the difficulties in modeling the three-dimensional configurations using two-dimensional methods. The two-dimensional calculations used a three-step procedure in which each of the three legs making the two 90-deg bends was a separate calculation. The experiment was recently analyzed with the TORT three-dimensional discrete ordinates radiation transport code, not only to compare the calculational results with the experimental results, but also to compare with results obtained from analyses in Japan using DOT3.5, MORSE, and ENSEMBLE, which is a three-dimensional discrete ordinates radiation transport code developed in Japan

Dosimeter incorporating radiophotoluminescent detectors for thermal neutrons and γ-rays in n-γ fields

Energy Technology Data Exchange (ETDEWEB)

Salem, Y.O. [Groupe RaMsEs, Institut Pluridisciplinaire Hubert Curien (IPHC), UMR 7178 CNRS/IN2P3, 23 rue du Loess, BP 28, F-67037 Strasbourg Cedex 2 (France); Nachab, A., E-mail: a.nachab@uca.ma [Département de physique, Faculté Poly-disciplinaire, Université Cadi Ayyad, Route Sidi Bouzid BP 4162, 46000 Safi (Morocco); Roy, C.; Nourreddine, A. [Groupe RaMsEs, Institut Pluridisciplinaire Hubert Curien (IPHC), UMR 7178 CNRS/IN2P3, 23 rue du Loess, BP 28, F-67037 Strasbourg Cedex 2 (France)

2016-10-15

We have developed a dosimeter associating different neutron converters with two radiophotoluminescent detectors to measure thermal neutrons and γ-rays in a mixed n-γ field. Tests show that the H{sup ∗}(10) and H{sub p}(10) responses to thermal neutrons and γ-rays are linear with detection limits lower than 0.4 mSv. The angular dependence of the dosimeter response is satisfactory and the influence of a phantom on the results is examined.

Two dimensional PMMA nanofluidic device fabricated by hot embossing and oxygen plasma assisted thermal bonding methods

Science.gov (United States)

Yin, Zhifu; Sun, Lei; Zou, Helin; Cheng, E.

2015-05-01

A method for obtaining a low-cost and high-replication precision two-dimensional (2D) nanofluidic device with a polymethyl methacrylate (PMMA) sheet is proposed. To improve the replication precision of the 2D PMMA nanochannels during the hot embossing process, the deformation of the PMMA sheet was analyzed by a numerical simulation method. The constants of the generalized Maxwell model used in the numerical simulation were calculated by experimental compressive creep curves based on previously established fitting formula. With optimized process parameters, 176 nm-wide and 180 nm-deep nanochannels were successfully replicated into the PMMA sheet with a replication precision of 98.2%. To thermal bond the 2D PMMA nanochannels with high bonding strength and low dimensional loss, the parameters of the oxygen plasma treatment and thermal bonding process were optimized. In order to measure the dimensional loss of 2D nanochannels after thermal bonding, a dimension loss evaluating method based on the nanoindentation experiments was proposed. According to the dimension loss evaluating method, the total dimensional loss of 2D nanochannels was 6 nm and 21 nm in width and depth, respectively. The tensile bonding strength of the 2D PMMA nanofluidic device was 0.57 MPa. The fluorescence images demonstrate that there was no blocking or leakage over the entire microchannels and nanochannels.

Characterization of the Ljubljana TRIGA thermal column neutron radiographic facility

International Nuclear Information System (INIS)

Nemec, T.; Rant, J.; Kristof, E.; Glumac, B.

1995-01-01

An extensive characterization of the neutron beam of the existing neutron radiographic facility in the thermal column of the Ljubljana Triga Mark II research reactor is in progress. Neutron beam characteristics are needed to determine the effect of various neutron and gamma radiation on the neutron radiographic image. Commercially available medical scintillator converter screens based on Gd dioxy sulphite as well as Gd metal neutron converters are used to record neutron radiographic image. Thermal, epithermal and fast neutron fluxes were measured using Au and In activation detectors and cadmium ratio is determined. Neutron beam flux profiles are measured by film densitometry and by Au activation detector wires. By exposing films shielded by boral or lead plates individual contributions of thermal, epithermal neutrons and gamma radiation are estimated by densitometric measurements. By recording images of neutron image quality indicators BPI (Beam Purity Indicator) and SI (Sensitivity Indicator) produced by Riso, standard neutron radiography image characteristic are established. In gamma dosimetric measurements thermoluminescent detectors (CaF 2 Mn) are used. (author)

A design study on hyper-thermal neutron irradiation field for neutron capture therapy at Kyoto University Reactor

International Nuclear Information System (INIS)

Sakurai, Y.; Kobayashi, T.

2000-01-01

A study about the installation of a hyper-thermal neutron converter to a clinical collimator was performed, as a series of the design study on a hyper-thermal neutron irradiation field at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. From the parametric-surveys by Monte Carlo calculation, it was confirmed that the practical irradiation field of hyper-thermal neutrons would be feasible by the modifications of the clinical collimator and the bismuth-layer structure. (author)

OPT-TWO: Calculation code for two-dimensional MOX fuel models in the optimum concentration distribution

International Nuclear Information System (INIS)

Sato, Shohei; Okuno, Hiroshi; Sakai, Tomohiro

2007-08-01

OPT-TWO is a calculation code which calculates the optimum concentration distribution, i.e., the most conservative concentration distribution in the aspect of nuclear criticality safety, of MOX (mixed uranium and plutonium oxide) fuels in the two-dimensional system. To achieve the optimum concentration distribution, we apply the principle of flattened fuel importance distribution with which the fuel system has the highest reactivity. Based on this principle, OPT-TWO takes the following 3 calculation steps iteratively to achieve the optimum concentration distribution with flattened fuel importance: (1) the forward and adjoint neutron fluxes, and the neutron multiplication factor, with TWOTRAN code which is a two-dimensional neutron transport code based on the SN method, (2) the fuel importance, and (3) the quantity of the transferring fuel. In OPT-TWO, the components of MOX fuel are MOX powder, uranium dioxide powder and additive. This report describes the content of the calculation, the computational method, and the installation method of the OPT-TWO, and also describes the application method of the criticality calculation of OPT-TWO. (author)

Improvement of neutron collimator design for thermal neutron radiography using Monte Carlo N-particle transport code version 5

International Nuclear Information System (INIS)

Thiagu Supramaniam

2007-01-01

The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

International Nuclear Information System (INIS)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

Multigroup or multipoint thermal neutron data preparation. Programme SIGMA

International Nuclear Information System (INIS)

Matausek, M.V.; Kunc, M.

1974-01-01

When calculating the space energy distribution of thermal neutrons in reactor lattices, in either the multigroup or the multipoint approximation, it is convenient to divide the problem into two independent parts. Firstly, for all material regions of the given reactor lattice cell, the group or the point values of cross sections, scattering kernel and the outer source of thermal neutrons are calculated by a data preparation programme. These quantities are then used as input, by the programme which solves multigroup or multipoint transport equations, to generate the space energy neutron spectra in the cell considered and to determine the related integral quantities, namely the different reaction rates. The present report deals with the first part of the problem. An algorithm for constructing a set of thermal neutron input data, to be used with the multigroup or multipoint version of the code MULTI /1,2,3/, is presented and the new version of the programme SIGMA /4/, written in FORTRAN IV for the CDC-3600 computer, is described. For a given reactor cell material, composed of a number of different isotopes, this programme calculates the group or the point values of the scattering macroscopic absorption cross section, macroscopic scattering cross section, kernel and the outer source of thermal neutrons. Numerous options are foreseen in the programme, concerning the energy variation of cross sections and a scattering kernel, concerning the weighting spectrum in multigroup scheme or the procedure for constructing the scattering matrix in the multipoint scheme and, finally, concerning the organization of output. The details of the calculational algorithm are presented in Section 2 of the paper. Section 3 contains the description of the programme and the instructions for its use (author)

A test-type hyper-thermal neutron generator for neutron capture therapy - estimation of neutron energy spectrum by simulation calculations and TOF experiments

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei

1999-01-01

In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments

Thermal conductivity of a two-dimensional phosphorene sheet: a comparative study with graphene.

Science.gov (United States)

Hong, Yang; Zhang, Jingchao; Huang, Xiaopeng; Zeng, Xiao Cheng

2015-11-28

A recently discovered two-dimensional (2D) layered material phosphorene has attracted considerable interest as a promising p-type semiconducting material. In this work, thermal conductivity (κ) of monolayer phosphorene is calculated using large-scale classical non-equilibrium molecular dynamics (NEMD) simulations. The predicted thermal conductivities for infinite length armchair and zigzag phosphorene sheets are 63.6 and 110.7 W m(-1) K(-1) respectively. The strong anisotropic thermal transport is attributed to the distinct atomic structures at altered chiral directions and direction-dependent group velocities. Thermal conductivities of 2D graphene sheets with the same dimensions are also computed for comparison. The extrapolated κ of the 2D graphene sheet are 1008.5(+37.6)(-37.6) and 1086.9(+59.1)(-59.1) W m(-1) K(-1) in the armchair and zigzag directions, respectively, which are an order of magnitude higher than those of phosphorene. The overall and decomposed phonon density of states (PDOS) are calculated in both structures to elucidate their thermal conductivity differences. In comparison with graphene, the vibrational frequencies that can be excited in phosphorene are severely limited. The temperature effect on the thermal conductivity of phosphorene and graphene sheets is investigated, which reveals a monotonic decreasing trend for both structures.

Applications of thermal neutron scattering

International Nuclear Information System (INIS)

Kostorz, G.

1978-01-01

Although in the past neutrons have been used quite frequently in the study of condensed matter, a more recent development has lead to applications of thermal neutron scattering in the investigation of more practical rather than purely academic problems. Physicists, chemists, materials scientists, biologists, and others have recognized and demonstrated that neutron scattering techniques can yield supplementary information which, in many cases, could not be obtained with other methods. The paper illustrates the use of neutron scattering in these areas of applied research. No attempt is made to present all the aspects of neutron scattering which can be found in textbooks. From the vast amount of experimental data, only a few examples are presented for the study of structure and atomic arrangement, ''extended'' structure, and dynamic phenomena in substances of current interest in applied research. (author)

Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

Science.gov (United States)

Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

2018-01-01

Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

Nondestructive elemental analysis of coins using accelerator-based thermal neutrons

International Nuclear Information System (INIS)

Khairi, F.Z.; Aksoy, A.; Al-Haddad, M.N.

2007-01-01

The accelerator-based thermal-neutrons activation analysis setup at KFUPM has an adequate thermal -neutron flux that can be advantageously used for the elemental analysis of a variety of samples including archeological ones. The thermal neutrons are derived from the moderation of fast neutrons from the D (d, n) He reaction which produces fast 2.5 MeV neutrons. A maximum thermals flux of about 2.5x10 n/m-s was achieved. For the purpose of determining the suitability of the set up for the analysis of contemporary and ancient coins, we carried out a feasibility study by irradiating a selected number of Saudi Arabian coins dating from 1958 to 1987 in the thermal-neutron flux. The induced gamma-ray activities were then counted using a HP-GMX detector coupled to a PC-based data acquisition and analysis system. The elements that were determined in the coins were copper (75%), nickel (around 25%) and manganese (<0.5%). Calibration curves were also established for these elements. The determined concentrations are in agreement with the data published by the Standard Catalogue of World Coins. (author)

A numerical method for two-dimensional anisotropic transport problem in cylindrical geometry

International Nuclear Information System (INIS)

Du Mingsheng; Feng Tiekai; Fu Lianxiang; Cao Changshu; Liu Yulan

1988-01-01

The authors deal with the triangular mesh-discontinuous finite element method for solving the time-dependent anisotropic neutron transport problem in two-dimensional cylindrical geometry. A prior estimate of the numerical solution is given. Stability is proved. The authors have computed a two dimensional anisotropic neutron transport problem and a Tungsten-Carbide critical assembly problem by using the numerical method. In comparision with DSN method and the experimental results obtained by others both at home and abroad, the method is satisfactory

Uncertainty and sensitivity analysis in the neutronic parameters generation for BWR and PWR coupled thermal-hydraulic–neutronic simulations

International Nuclear Information System (INIS)

Ánchel, F.; Barrachina, T.; Miró, R.; Verdú, G.; Juanas, J.; Macián-Juan, R.

2012-01-01

Highlights: ► Best-estimate codes are affected by the uncertainty in the methods and the models. ► Influence of the uncertainty in the macroscopic cross-sections in a BWR and PWR RIA accidents analysis. ► The fast diffusion coefficient, the scattering cross section and both fission cross sections are the most influential factors. ► The absorption cross sections very little influence. ► Using a normal pdf the results are more “conservative” comparing the power peak reached with uncertainty quantified with a uniform pdf. - Abstract: The Best Estimate analysis consists of a coupled thermal-hydraulic and neutronic description of the nuclear system's behavior; uncertainties from both aspects should be included and jointly propagated. This paper presents a study of the influence of the uncertainty in the macroscopic neutronic information that describes a three-dimensional core model on the most relevant results of the simulation of a Reactivity Induced Accident (RIA). The analyses of a BWR-RIA and a PWR-RIA have been carried out with a three-dimensional thermal-hydraulic and neutronic model for the coupled system TRACE-PARCS and RELAP-PARCS. The cross section information has been generated by the SIMTAB methodology based on the joint use of CASMO-SIMULATE. The statistically based methodology performs a Monte-Carlo kind of sampling of the uncertainty in the macroscopic cross sections. The size of the sampling is determined by the characteristics of the tolerance intervals by applying the Noether–Wilks formulas. A number of simulations equal to the sample size have been carried out in which the cross sections used by PARCS are directly modified with uncertainty, and non-parametric statistical methods are applied to the resulting sample of the values of the output variables to determine their intervals of tolerance.

Spherical harmonics solutions of multi-dimensional neutron transport equation by finite Fourier transformation

International Nuclear Information System (INIS)

Kobayashi, Keisuke

1977-01-01

A method of solution of a monoenergetic neutron transport equation in P sub(L) approximation is presented for x-y and x-y-z geometries using the finite Fourier transformation. A reactor system is assumed to consist of multiregions in each of which the nuclear cross sections are spatially constant. Since the unknown functions of this method are the spherical harmonics components of the neutron angular flux at the material boundaries alone, the three- and two-dimensional equations are reduced to two- and one-dimensional equations, respectively. The present approach therefore gives fewer unknowns than in the usual series expansion method or in the finite difference method. Some numerical examples are shown for the criticality problem. (auth.)

Thermal neutron diffusion parameters dependent on the flux energy distribution in finite hydrogenous media

International Nuclear Information System (INIS)

Drozdowicz, K.

1999-01-01

cooling function, is presented. On this basis the diffusion cooling coefficient is calculated for two hydrogenous materials (Plexiglas, polyethylene). A method is presented to calculate the thermal neutron energy-averaged (stationary and pulsed) parameters of a hydrogenous medium for which the energy dependence of the scattering cross section is unknown. New effective scattering cross sections of hydrogen are determined for this purpose. (author)

The single-collision thermalization approximation for application to cold neutron moderation problems

International Nuclear Information System (INIS)

Ritenour, R.L.

1989-01-01

The single collision thermalization (SCT) approximation models the thermalization process by assuming that neutrons attain a thermalized distribution with only a single collision within the moderating material, independent of the neutron's incident energy. The physical intuition on which this approximation is based is that the salient properties of neutron thermalization are accounted for in the first collision, and the effects of subsequent collisions tend to average out statistically. The independence of the neutron incident and outscattering energy leads to variable separability in the scattering kernel and, thus, significant simplification of the neutron thermalization problem. The approximation also addresses detailed balance and neutron conservation concerns. All of the tests performed on the SCT approximation yielded excellent results. The significance of the SCT approximation is that it greatly simplifies thermalization calculations for CNS design. Preliminary investigations with cases involving strong absorbers also indicates that this approximation may have broader applicability, as in the upgrading of the thermalization codes

Intercomparison of personnel dosimetry for thermal neutron dose equivalent in neutron and gamma-ray mixed fields

International Nuclear Information System (INIS)

Ogawa, Yoshihiro

1985-01-01

In order to consider the problems concerned with personnel dosimetry using film badges and TLDs, an intercomparison of personnel dosimetry, especially dose equivalent responses of personnel dosimeters to thermal neutron, was carried out in five different neutron and gamma-ray mixed fields at KUR and UTR-KINKI from the practical point of view. For the estimation of thermal neutron dose equivalent, it may be concluded that each personnel dosimeter has good performances in the precision, that is, the standard deviations in the measured values by individual dosimeter were within 24 %, and the dose equivalent responses to thermal neutron were almost independent on cadmium ratio and gamma-ray contamination. However, the relative thermal neutron dose equivalent of individual dosimeter normalized to the ICRP recommended value varied considerably and a difference of about 4 times was observed among the dosimeters. From the results obtained, it is suggested that the standardization of calibration factors and procedures is required from the practical point of radiation protection and safety. (author)

MCNP Simulations of Measurement of Insulation Compaction in the Cryogenic Rocket Fuel Tanks at Kennedy Space Center by Fast/Thermal Neutron Techniques

Science.gov (United States)

Livingston, R. A.; Schweitzer, J. S.; Parsons, A. M.; Arens, E. E.

2010-01-01

MCNP simulations have been run to evaluate the feasibility of using a combination of fast and thermal neutrons as a nondestructive method to measure of the compaction of the perlite insulation in the liquid hydrogen and oxygen cryogenic storage tanks at John F. Kennedy Space Center (KSC). Perlite is a feldspathic volcanic rock made up of the major elements Si, AI, Na, K and 0 along with some water. When heated it expands from four to twenty times its original volume which makes it very useful for thermal insulation. The cryogenic tanks at Kennedy Space Center are spherical with outer diameters of 69-70 feet and lined with a layer of expanded perlite with thicknesses on the order of 120 cm. There is evidence that some of the perlite has compacted over time since the tanks were built 1965, affecting the thermal properties and possibly also the structural integrity of the tanks. With commercially available portable neutron generators it is possible to produce simultaneously fluxes of neutrons in two energy ranges: fast (14 Me V) and thermal (25 me V). The two energy ranges produce complementary information. Fast neutrons produce gamma rays by inelastic scattering, which is sensitive to Fe and O. Thermal neutrons produce gamma rays by prompt gamma neutron activation (PGNA) and this is sensitive to Si, Al, Na, K and H. The compaction of the perlite can be measured by the change in gamma ray signal strength which is proportional to the atomic number densities of the constituent elements. The MCNP simulations were made to determine the magnitude of this change. The tank wall was approximated by a I-dimensional slab geometry with an 11/16" outer carbon steel wall, an inner stainless wall and 120 cm thick perlite zone. Runs were made for cases with expanded perlite, compacted perlite or with various void fractions. Runs were also made to simulate the effect of adding a moderator. Tallies were made for decay-time analysis from t=0 to 10 ms; total detected gamma

RBE of thermal neutrons for induction of chromosome aberrations in human lymphocytes.

Science.gov (United States)

Schmid, E; Wagner, F M; Canella, L; Romm, H; Schmid, T E

2013-03-01

The induction of chromosome aberrations in human lymphocytes irradiated in vitro with slow neutrons was examined to assess the maximum low-dose RBE (RBE(M)) relative to (60)Co γ-rays. For the blood irradiations, cold neutron beam available at the prompt gamma activation analysis facility at the Munich research reactor FRM II was used. The given flux of cold neutrons can be converted into a thermally equivalent one. Since blood was taken from the same donor whose blood had been used for previous irradiation experiments using widely varying neutron energies, the greatest possible accuracy was available for such an estimation of the RBE(M) avoiding the inter-individual variations or differences in methodology usually associated with inter-laboratory comparisons. The magnitude of the coefficient α of the linear dose-response relationship (α = 0.400 ± 0.018 Gy(-1)) and the derived RBE(M) of 36.4 ± 13.3 obtained for the production of dicentrics by thermal neutrons confirm our earlier observations of a strong decrease in α and RBE(M) with decreasing neutron energy lower than 0.385 MeV (RBE(M) = 94.4 ± 38.9). The magnitude of the presently estimated RBE(M) of thermal neutrons is-with some restrictions-not significantly different to previously reported RBE(M) values of two laboratories.

Maximum neutron flux in thermal reactors

International Nuclear Information System (INIS)

Strugar, P.V.

1968-12-01

Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples

Applications of thermal neutron scattering in biology, biochemistry and biophysics

International Nuclear Information System (INIS)

Worcester, D.L.

1977-01-01

Biological applications of thermal neutron scattering have increased rapidly in recent years. The following categories of biological research with thermal neutron scattering are presently identified: crystallography of biological molecules; neutron small-angle scattering of biological molecules in solution (these studies have already included numerous measurements of proteins, lippoproteins, viruses, ribosomal subunits and chromatin subunit particles); neutron small-angle diffraction and scattering from biological membranes and membrane components; and neutron quasielastic and inelastic scattering studies of the dynamic properties of biological molecules and materials. (author)

New thermal neutron calibration channel at LNMRI/IRD

International Nuclear Information System (INIS)

Astuto, A.; Lopes, R.T.; Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W.

2015-01-01

A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four 241 Am-Be sources with 596 GBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence (less than 1%) in the central chamber. (author)

An analytical discrete ordinates solution for a nodal model of a two-dimensional neutron transport problem

International Nuclear Information System (INIS)

Filho, J. F. P.; Barichello, L. B.

2013-01-01

In this work, an analytical discrete ordinates method is used to solve a nodal formulation of a neutron transport problem in x, y-geometry. The proposed approach leads to an important reduction in the order of the associated eigenvalue systems, when combined with the classical level symmetric quadrature scheme. Auxiliary equations are proposed, as usually required for nodal methods, to express the unknown fluxes at the boundary introduced as additional unknowns in the integrated equations. Numerical results, for the problem defined by a two-dimensional region with a spatially constant and isotropically emitting source, are presented and compared with those available in the literature. (authors)

Using TRIGA Mark II research reactor for irradiation with thermal neutrons

Energy Technology Data Exchange (ETDEWEB)

Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

2015-03-15

Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

On the use of silicon as thermal neutron filter

International Nuclear Information System (INIS)

Adib, M.; Habib, N.; Ashry, A.; Fathalla, M.

2003-01-01

A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 μeV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given

On the use of silicon as thermal neutron filter

Energy Technology Data Exchange (ETDEWEB)

Adib, M.; Habib, N.; Ashry, A.; Fathalla, M. E-mail: mohamedfathalla@hotmail.com

2003-12-01

A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 {mu}eV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given.

Boron neutron capture therapy (BNCT). Recent aspect, a change from thermal neutron to epithermal neutron beam and a new protocol

International Nuclear Information System (INIS)

Nakagawa, Yoshinobu

1999-01-01

Since 1968, One-hundred seventy three patients with glioblastoma (n=81), anaplastic astrocytoma (n=44), low grade astrocytoma (n=16) or other types of tumor (n=32) were treated by boron-neutron capture therapy (BNCT) using a combination of thermal neutron and BSH in 5 reactors (HTR n=13, JRR-3 n=1, MuITR n=98, KUR n=28, JRR-2 n=33). Out of 101 patients with glioma treated by BNCT under the recent protocol, 33 (10 glioblastoma, 14 anaplastic astrocytoma, 9 low grade astrocytoma) patients lived or have lived longer than 3 years. Nine of these 33 lived or have lived longer than 10 years. According to the retrospective analysis, the important factors related to the clinical results were tumor dose radiation dose and maximum radiation dose in thermal brain cortex. The result was not satisfied as it was expected. Then, we decided to introduce mixed beams which contain thermal neutron and epithermal neutron beams. KUR was reconstructed in 1996 and developed to be available to use mixed beams. Following the shutdown of the JRR-2, JRR-4 was renewed for medical use in 1998. Both reactors have capacity to yield thermal neutron beam, epithermal neutron beam and mixed beams. The development of the neutron source lead us to make a new protocol. (author)

Two-dimensional finite element neutron diffusion analysis using hierarchic shape functions

International Nuclear Information System (INIS)

Carpenter, D.C.

1997-01-01

Recent advances have been made in the use of p-type finite element method (FEM) for structural and fluid dynamics problems that hold promise for reactor physics problems. These advances include using hierarchic shape functions, element-by-element iterative solvers and more powerful mapping techniques. Use of the hierarchic shape functions allows greater flexibility and efficiency in implementing energy-dependent flux expansions and incorporating localized refinement of the solution space. The irregular matrices generated by the p-type FEM can be solved efficiently using element-by-element conjugate gradient iterative solvers. These solvers do not require storage of either the global or local stiffness matrices and can be highly vectorized. Mapping techniques based on blending function interpolation allow exact representation of curved boundaries using coarse element grids. These features were implemented in a developmental two-dimensional neutron diffusion program based on the use of hierarchic shape functions (FEM2DH). Several aspects in the effective use of p-type analysis were explored. Two choices of elemental preconditioning were examined--the proper selection of the polynomial shape functions and the proper number of functions to use. Of the five shape function polynomials tested, the integral Legendre functions were the most effective. The serendipity set of functions is preferable over the full tensor product set. Two global preconditioners were also examined--simple diagonal and incomplete Cholesky. The full effectiveness of the finite element methodology was demonstrated on a two-region, two-group cylindrical problem but solved in the x-y coordinate space, using a non-structured element grid. The exact, analytic eigenvalue solution was achieved with FEM2DH using various combinations of element grids and flux expansions

Analysis of the Photoneutron Yield and Thermal Neutron Flux in an Unreflected Electron Accelerator-Driven Neutron Source

International Nuclear Information System (INIS)

Dale, Gregory E.; Gahl, John M.

2005-01-01

There are several potential uses for a high-flux thermal neutron source in both industrial and clinical applications. The viable commercial implementation of these applications requires a low-cost, high-flux thermal neutron generator suitable for installation in industrial and clinical environments. This paper describes the Monte Carlo for N-Particle modeling results of a high-flux thermal neutron source driven with an electron accelerator. An electron linear accelerator (linac), fitted with a standard X-ray converter, can produce high neutron yields in materials with low photonuclear threshold energies, such as D and 9 Be. Results indicate that a 10-MeV, 10-kW electron linac can produce on the order of 10 12 n/s in a heavy water photoneutron target. The thermal neutron flux in an unreflected heavy water target is calculated to be on the order of 10 10 n.cm -2 .s. The sensitivity of these answers to heavy water purity is also investigated, specifically the dilution of heavy water with light water. It is shown that the peak thermal neutron flux is not adversely effected by dilution up to a light water weight fraction of 35%

Thermal neutron capture gamma-rays

International Nuclear Information System (INIS)

Tuli, J.K.

1983-01-01

The energy and intensity of gamma rays as seen in thermal neutron capture are presented. Only those (n,α), E = thermal, reactions for which the residual nucleus mass number is greater than or equal to 45 are included. These correspond to evaluations published in Nuclear Data Sheets. The publication source data are contained in the Evaluated Nuclear Structure Data File (ENSDF). The data presented here do not involve any additional evaluation. Appendix I lists all the residual nuclides for which the data are included here. Appendix II gives a cumulated index to A-chain evaluations including the year of publication. The capture gamma ray data are given in two tables - the Table 1 is the list of all gamma rays seen in (n,#betta#) reaction given in the order of increasing energy; the Table II lists the gamma rays according to the nuclide

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II forced feed reflood tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1987-01-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase of a PWR-LOCA. It was revealed in the previous Slab Core Test Facility (SCTF) Core-II test results that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. In order to separately evaluate the effect of the radial power (Q) distribution itself and the effect of the radial temperature (T) distribution, four tests were performed with steep Q and T, flat Q and T, steep Q and flat T, and flat Q and steep T. Based on the test results, it was concluded that the radial temperature distribution which accompanied the radial power distribution was the dominant factor of the two-dimensional thermal-hydraulic behavior in the core during the initial period. Selected data from these four tests are also presented in this report. Some data from Test S2-12 (steep Q, T) were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

Bibliography for thermal neutron scattering

International Nuclear Information System (INIS)

Sakamoto, M.; Chihara, J.; Nakahara, Y.; Kadotani, H.; Sekiya, T.

1976-12-01

It contains bibliographical references to measurements, calculations, reviews and basic studies on thermal neutron scatterings and dynamical properties of condensed matter. About 2,700 documents up to the end of 1975 are covered. (auth.)

Development of a hybrid MSGC detector for thermal neutron imaging with a MHz data acquisition and histogramming system

CERN Document Server

Gebauer, B; Richter, G; Levchanovsky, F V; Nikiforov, A

2001-01-01

For thermal neutron imaging at the next generation of high-flux pulsed neutron sources a large area and fourfold segmented, hybrid, low-pressure, two-dimensional position sensitive, microstrip gas chamber detector, fabricated in a multilayer technology on glass substrates, is presently being developed, which utilizes a thin composite sup 1 sup 5 sup 7 Gd/CsI neutron converter. The present article focusses on the readout scheme and the data acquisition (DAQ) system. For position encoding, interpolating and fast multihit delay line based electronics is applied with up to eightfold sub-segmentation per geometrical detector segment. All signals, i.e. position, time-of-flight and pulse-height signals, are fed into deadtime-less 8-channel multihit TDC chips with 120 ps LSB via constant fraction and time-over-threshold discriminators, respectively. The multihit capability is utilized to raise the count rate limit in combination with a sum check algorithm for disentangling pulses from different events. The first vers...

Magneto–Thermal Evolution of Neutron Stars with Emphasis to ...

Indian Academy of Sciences (India)

The magnetic and thermal evolution of neutron stars is a very complex process with many non-linear interactions. For a decent understanding of neutron star physics, these evolutions cannot be considered isolated. A brief overview is presented, which describes the main magneto–thermal interactions that determine the fate ...

Study and development of new dosemeters for thermal neutrons; Estudio y desarrollo de nuevos dosimetros para neutrones termicos

Energy Technology Data Exchange (ETDEWEB)

Urena N, F

1998-12-31

An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation ({phi} = 5 x 10{sup 7} n/cm{sup 2} -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence {Phi} = {phi} {Delta}t ; where {Delta}t = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

Neutronics methods for thermal radiative transfer

International Nuclear Information System (INIS)

Larsen, E.W.

1988-01-01

The equations of thermal radiative transfer are time discretized in a semi-implicit manner, yielding a linear transport problem for each time step. The governing equation in this problem has the form of a neutron transport equation with fission but no scattering. Numerical methods are described, whose origins lie in neutron transport, and that have been successfully adapted to this new problem. Acceleration methods that have been developed specifically for the radiative transfer problem, but may have generalizations applicable in neutronics problems, are also discussed

Assessment of the TASS 1-D neutronics model for the westinghouse and ABB-CE type PWR reactivity induced transients

International Nuclear Information System (INIS)

Choi, J.D.; Yoon, H.Y.; Um, K.S.; Kim, H.C.; Sim, S.K.

1997-01-01

Best estimate transient analysis code, TASS, has been developed for the normal and transient simulation of the Westinghouse and ABB-CE type PWRs. TASS thermal hydraulic model is based on the non-homogeneous, non-equilibrium two-phase continuity, energy and mixture momentum equations with constitutive relations for closure. Core neutronics model employs both the point kinetics and one-dimensional neutron diffusion model. Semi-implicit numerical scheme is used to solve the discretized finite difference equations. TASS one dimensional neutronics core model has been assessed through the reactivity induced transient analyses for the KORI-3, three loop Westinghouse PWR, and Younggwang-3 (YGN-3), two-loop ABB-CE PWR, nuclear power plants currently operating in Korea. The assessment showed that the TASS one dimensional neutronics core model can be applied for the Westinghouse and ABB-CE type PWRs to gain thermal margin which is necessary for a potential use of the high fuel burnup, extended fuel cycle, power upgrading and for the plant life extension

Influence of the effective mass of water molecule on thermal neutron scattering

International Nuclear Information System (INIS)

Markovic, M.

1981-01-01

The influence of the effective water molecule mass on the thermal neutron scattering on the nucleus of the hydrogen atom has been investigated. Besides the actual water molecule mass (M = 18) the investigations have been carried out with its two effective values (M1 = 16 and M2 = 20). The differential and total cross sections have been calculated for the incident thermal neutron energy E o = 1 eV. Investigation results show different prominence of the quantum effects and for M2 the appearance of peaks in the quasielastic scattering. (author)

Thickness optimization of various moderator materials for maximization of thermal neutron fluence

International Nuclear Information System (INIS)

Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

2015-01-01

Plasma focus device is widely being used as pulsed neutron source for variety of applications. Measurements of neutron yield by largely preferred Helium-3 proportional counter and Silver activation counter are mainly sensitive to thermal neutrons and are typically used with a neutron moderator. Thermalization of neutron is based on scattering reaction and hydrogenous materials are the best thermalizing medium. The efficiency of aforementioned neutron detectors is considerably affected by physical and geometrical properties of thermalizing medium i.e. moderator material, its thickness and shape. In view of the same, simulations have been performed to explore the effective utilization of Polyethylene, Perspex and Light water as moderating mediums for cylindrical and spherical geometry. In this study, estimated thermal fluence value up to 0.5 eV has been considered as the benchmark factor for comparing efficient thermalization by specific material, its thickness and shape. In either of the shapes being cylindrical or spherical, use of Polyethylene as moderating medium has resulted in minimum optimum thickness along with highest thermal fluence. (author)

A Monte Carlo Green's function method for three-dimensional neutron transport

International Nuclear Information System (INIS)

Gamino, R.G.; Brown, F.B.; Mendelson, M.R.

1992-01-01

This paper describes a Monte Carlo transport kernel capability, which has recently been incorporated into the RACER continuous-energy Monte Carlo code. The kernels represent a Green's function method for neutron transport from a fixed-source volume out to a particular volume of interest. This method is very powerful transport technique. Also, since kernels are evaluated numerically by Monte Carlo, the problem geometry can be arbitrarily complex, yet exact. This method is intended for problems where an ex-core neutron response must be determined for a variety of reactor conditions. Two examples are ex-core neutron detector response and vessel critical weld fast flux. The response is expressed in terms of neutron transport kernels weighted by a core fission source distribution. In these types of calculations, the response must be computed for hundreds of source distributions, but the kernels only need to be calculated once. The advance described in this paper is that the kernels are generated with a highly accurate three-dimensional Monte Carlo transport calculation instead of an approximate method such as line-of-sight attenuation theory or a synthesized three-dimensional discrete ordinates solution

Two-dimensional turbulent convection

Science.gov (United States)

Mazzino, Andrea

2017-11-01

We present an overview of the most relevant, and sometimes contrasting, theoretical approaches to Rayleigh-Taylor and mean-gradient-forced Rayleigh-Bénard two-dimensional turbulence together with numerical and experimental evidences for their support. The main aim of this overview is to emphasize that, despite the different character of these two systems, especially in relation to their steadiness/unsteadiness, turbulent fluctuations are well described by the same scaling relationships originated from the Bolgiano balance. The latter states that inertial terms and buoyancy terms balance at small scales giving rise to an inverse kinetic energy cascade. The main difference with respect to the inverse energy cascade in hydrodynamic turbulence [R. H. Kraichnan, "Inertial ranges in two-dimensional turbulence," Phys. Fluids 10, 1417 (1967)] is that the rate of cascade of kinetic energy here is not constant along the inertial range of scales. Thanks to the absence of physical boundaries, the two systems here investigated turned out to be a natural physical realization of the Kraichnan scaling regime hitherto associated with the elusive "ultimate state of thermal convection" [R. H. Kraichnan, "Turbulent thermal convection at arbitrary Prandtl number," Phys. Fluids 5, 1374-1389 (1962)].

Neutron radiography

International Nuclear Information System (INIS)

Hiraoka, Eiichi

1988-01-01

The thermal neutron absorption coefficient is essentially different from the X-ray absorption coefficient. Each substance has a characteristic absorption coefficient regardless of its density. Neutron deams have the following features: (1) neutrons are not transmitted efficiently by low molecular weight substances, (2) they are transmitted efficiently by heavy metals, and (3) the transmittance differs among isotopes. Thus, neutron beams are suitable for cheking for foreign matters in heavy metals and testing of composites consisting of both heavy and light materials. A neutron source generates fast neutrons, which should be converted into thermal neutrons by reducing their energy. Major neutron souces include nuclear reactors, radioisotopes and particle accelerators. Photographic films and television systems are mainly used to observe neutron transmission images. Computers are employed for image processing, computerized tomography and three-dimensional analysis. The major applications of neutron radiography include inspection of neclear fuel; evaluation of material for airplane; observation of fuel in the engine and oil in the hydraulic systems in airplanes; testing of composite materials; etc. (Nogami, K.)

Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

International Nuclear Information System (INIS)

Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.

2017-01-01

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

Energy Technology Data Exchange (ETDEWEB)

Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

2017-01-15

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

The measurements of thermal neutron flux distribution in a paraffin

Indian Academy of Sciences (India)

The term `thermal flux' implies a Maxwellian distribution of velocity and energy corresponding to the most probable velocity of 2200 ms-1 at 293.4 K. In order to measure the thermal neutron flux density, the foil activation method was used. Thermal neutron flux determination in paraffin phantom by counting the emitted rays of ...

Determination of the thermal and epithermal neutron sensitivities of an LBO chamber

Energy Technology Data Exchange (ETDEWEB)

Endo, Satoru; Kotani, Kei; Kajimoto, Tsuyoshi; Tanaka, Kenichi [Hiroshima University, Quantum Energy Applications, Graduate School of Engineering, Higashi-Hiroshima (Japan); Sato, Hitoshi; Nakajima, Erika [Ibaraki Prefectural University of Health Science, Radiological Sciences, Ibaraki (Japan); Shimazaki, Takuto [Hiroshima University, Quantum Energy Applications, Graduate School of Engineering, Higashi-Hiroshima (Japan); Delta Kogyo Co., Ltd., Hiroshima (Japan); Suda, Mitsuru; Hamano, Tsuyoshi [National Institute of Radiological Sciences, Chiba-Shi, Chiba (Japan); Hoshi, Masaharu [Hiroshima University, Institute for Peace Science, Hiroshima (Japan)

2017-08-15

An LBO (Li{sub 2}B{sub 4}O{sub 7}) walled ionization chamber was designed to monitor the epithermal neutron fluence in boron neutron capture therapy clinical irradiation. The thermal and epithermal neutron sensitivities of the device were evaluated using accelerator neutrons from the {sup 9}Be(d, n) reaction at a deuteron energy of 4 MeV (4 MeV d-Be neutrons). The response of the chamber in terms of the electric charge induced in the LBO chamber was compared with the thermal and epithermal neutron fluences measured using the gold-foil activation method. The thermal and epithermal neutron sensitivities obtained were expressed in units of pC cm{sup 2}, i.e., from the chamber response divided by neutron fluence (cm{sup -2}). The measured LBO chamber sensitivities were 2.23 x 10{sup -7} ± 0.34 x 10{sup -7} (pC cm{sup 2}) for thermal neutrons and 2.00 x 10{sup -5} ± 0.12 x 10{sup -5} (pC cm{sup 2}) for epithermal neutrons. This shows that the LBO chamber is sufficiently sensitive to epithermal neutrons to be useful for epithermal neutron monitoring in BNCT irradiation. (orig.)

Two-dimensional flexible nanoelectronics

Science.gov (United States)

Akinwande, Deji; Petrone, Nicholas; Hone, James

2014-12-01

2014/2015 represents the tenth anniversary of modern graphene research. Over this decade, graphene has proven to be attractive for thin-film transistors owing to its remarkable electronic, optical, mechanical and thermal properties. Even its major drawback--zero bandgap--has resulted in something positive: a resurgence of interest in two-dimensional semiconductors, such as dichalcogenides and buckled nanomaterials with sizeable bandgaps. With the discovery of hexagonal boron nitride as an ideal dielectric, the materials are now in place to advance integrated flexible nanoelectronics, which uniquely take advantage of the unmatched portfolio of properties of two-dimensional crystals, beyond the capability of conventional thin films for ubiquitous flexible systems.

Cross-section of single-crystal materials used as thermal neutron filters

International Nuclear Information System (INIS)

Adib, M.

2005-01-01

Transmission properties of several single crystal materials important for neutron scattering instrumentation are presented. A computer codes are developed which permit the calculation of thermal diffuse and Bragg-scattering cross-sections of silicon., and sapphire as a function of material's constants, temperature and neutron energy, E, in the range 0.1 MeV .A discussion of the use of their single-crystal as a thermal neutron filter in terms of the optimum crystal thickness, mosaic spread, temperature, cutting plane and tuning for efficient transmission of thermal-reactor neutrons is given

Three-dimensional coupled kinetics/thermal- hydraulic benchmark TRIGA experiments

International Nuclear Information System (INIS)

Feltus, Madeline Anne; Miller, William Scott

2000-01-01

This research project provides separate effects tests in order to benchmark neutron kinetics models coupled with thermal-hydraulic (T/H) models used in best-estimate codes such as the Nuclear Regulatory Commission's (NRC) RELAP and TRAC code series and industrial codes such as RETRAN. Before this research project was initiated, no adequate experimental data existed for reactivity initiated transients that could be used to assess coupled three-dimensional (3D) kinetics and 3D T/H codes which have been, or are being developed around the world. Using various Test Reactor Isotope General Atomic (TRIGA) reactor core configurations at the Penn State Breazeale Reactor (PSBR), it is possible to determine the level of neutronics modeling required to describe kinetics and T/H feedback interactions. This research demonstrates that the small compact PSBR TRIGA core does not necessarily behave as a point kinetics reactor, but that this TRIGA can provide actual test results for 3D kinetics code benchmark efforts. This research focused on developing in-reactor tests that exhibited 3D neutronics effects coupled with 3D T/H feedback. A variety of pulses were used to evaluate the level of kinetics modeling needed for prompt temperature feedback in the fuel. Ramps and square waves were used to evaluate the detail of modeling needed for the delayed T/H feedback of the coolant. A stepped ramp was performed to evaluate and verify the derived thermal constants for the specific PSBR TRIGA core loading pattern. As part of the analytical benchmark research, the STAR 3D kinetics code (, STAR: Space and time analysis of reactors, Version 5, Level 3, Users Guide, Yankee Atomic Electric Company, YEAC 1758, Bolton, MA) was used to model the transient experiments. The STAR models were coupled with the one-dimensional (1D) WIGL and LRA and 3D COBRA (, COBRA IIIC: A digital computer program for steady-state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements, Battelle

Characteristics of thermal neutron calibration fields using a graphite pile

International Nuclear Information System (INIS)

Uchita, Yoshiaki; Saegusa, Jun; Kajimoto, Yoichi; Tanimura, Yoshihiko; Shimizu, Shigeru; Yoshizawa, Michio

2005-03-01

The Facility of Radiation Standards of Japan Atomic Energy Research Institute is equipped with thermal neutron fields for calibrating area and personal neutron dosemeters. The fields use moderated neutrons leaked from a graphite pile in which radionuclide sources are placed. In January 2003, we have renewed the pile with some modifications in its size. In accordance with the renewal, we measured and calculated thermal neutron fluence rates, neutron energy distributions and angular distributions of the fields. The thermal neutron fluence rates of the ''inside-pile fields'' and the outside-pile fields'' were determined by the gold foil activation method. The neutron energy distributions of the outside-pile fields were also measured with the Bonner multi-sphere spectrometer system. The contributions of epithermal and fast neutrons to the total dose-equivalents were 9% in the southern outside-pile field and 12% in the western outside-pile field. The personal dose-equivalents, H p,slab (10, α), in the outside-pile fields are evaluated by considering the calculated angular distributions of incoming neutrons. The H p,slab (10, α) was found to be about 40% higher than the value in assuming the unidirectional neutron between the pile and the test point. (author)

HEXAGA-II-120, -60, -30 two-dimensional multi-group neutron diffusion programmes for a uniform triangular mesh with arbitrary group scattering

International Nuclear Information System (INIS)

Woznicki, Z.

1979-06-01

This report presents the AGA two-sweep iterative methods belonging to the family of factorization techniques in their practical application in the HEXAGA-II two-dimensional programme to obtain the numerical solution to the multi-group, time-independent, (real and/or adjoint) neutron diffusion equations for a fine uniform triangular mesh. An arbitrary group scattering model is permitted. The report written for the users provides the description of input and output. The use of HEXAGA-II is illustrated by two sample reactor problems. (orig.) [de

Measurement of thermal neutron capture cross section

International Nuclear Information System (INIS)

Huang Xiaolong; Han Xiaogang; Yu Weixiang; Lu Hanlin; Zhao Wenrong

2001-01-01

The thermal neutron capture cross sections of 71 Ga(n, γ) 72 Ga, 94 Zr(n, γ) 95 Zr and 191 Ir(n, γ) 192 Ir m1+g,m2 reactions were measured by using activation method and compared with other measured data. Meanwhile the half-life of 72 Ga was also measured. The samples were irradiated with the neutron in the thermal column of heavy water reactor of China Institute of Atomic Energy. The activities of the reaction products were measured by well-calibrated Ge(Li) detector

Quantitative one-dimensional thermal-wave cavity measurements of fluid thermophysical properties through equivalence studies with three-dimensional geometries

International Nuclear Information System (INIS)

Matvienko, Anna; Mandelis, Andreas

2006-01-01

The thermal-wave field in a photopyroelectric thermal-wave cavity was calculated with two theoretical approaches: a computationally straightforward, conventional, one-dimensional approach and a three-dimensional experimentally more realistic approach. The calculations show that the dimensionality of the thermal-wave field in the cavity depends on the lateral heat transfer boundary conditions and the relation between the beam size of the laser impinging on the thermal-wave generating metallic film and the diameter of the film itself. The theoretical calculations and the experimental data on the photopyroelectric signal in the cavity were compared. The study resulted in identifying ranges of heat transfer rates, beam sizes, and cavity radii for which accurate quantitative measurements of the thermal diffusivity of intracavity fluids can be made within the far simpler, but only approximate, one-dimensional approach conventionally adopted by users of thermal-wave cavities. It was shown that the major parameters affecting the dimensionality of thermal-wave cavities are the laser beam spot size and the Biot number of the medium comprising the sidewalls of the (cylindrical) cavity

A novel detector assembly for detecting thermal neutrons, fast neutrons and gamma rays

Energy Technology Data Exchange (ETDEWEB)

Cester, D., E-mail: davide.cester@gmail.com [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); Lunardon, M.; Moretto, S. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); INFN Sezione di Padova, Via Marzolo 8, I-35131 Padova (Italy); Nebbia, G. [INFN Sezione di Padova, Via Marzolo 8, I-35131 Padova (Italy); Pino, F. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); Sajo-Bohus, L. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy); Laboratorio de Fisica Nuclear, Universidad Simon Bolivar, Apartado 89000, 1080 A Caracas (Venezuela, Bolivarian Republic of); Stevanato, L.; Bonesso, I.; Turato, F. [Dipartimento di Fisica ed Astronomia dell' Università di Padova, Via Marzolo 8, I-35131 Padova (Italy)

2016-09-11

A new composite detector has been developed by combining two different commercial scintillators. The device has the capability to detect gamma rays as well as thermal and fast neutrons; the signal discrimination between the three types is performed on-line by means of waveform digitizers and PSD algorithms. This work describes the assembled detector and its discrimination performance to be employed in the applied field.

A novel detector assembly for detecting thermal neutrons, fast neutrons and gamma rays

International Nuclear Information System (INIS)

Cester, D.; Lunardon, M.; Moretto, S.; Nebbia, G.; Pino, F.; Sajo-Bohus, L.; Stevanato, L.; Bonesso, I.; Turato, F.

2016-01-01

A new composite detector has been developed by combining two different commercial scintillators. The device has the capability to detect gamma rays as well as thermal and fast neutrons; the signal discrimination between the three types is performed on-line by means of waveform digitizers and PSD algorithms. This work describes the assembled detector and its discrimination performance to be employed in the applied field.

Two-dimensional analytical solution for nodal calculation of nuclear reactors

International Nuclear Information System (INIS)

Silva, Adilson C.; Pessoa, Paulo O.; Silva, Fernando C.; Martinez, Aquilino S.

2017-01-01

Highlights: • A proposal for a coarse mesh nodal method is presented. • The proposal uses the analytical solution of the two-dimensional neutrons diffusion equation. • The solution is performed homogeneous nodes with dimensions of the fuel assembly. • The solution uses four average fluxes on the node surfaces as boundary conditions. • The results show good accuracy and efficiency. - Abstract: In this paper, the two-dimensional (2D) neutron diffusion equation is analytically solved for two energy groups (2G). The spatial domain of reactor core is divided into a set of nodes with uniform nuclear parameters. To determine iteratively the multiplication factor and the neutron flux in the reactor we combine the analytical solution of the neutron diffusion equation with an iterative method known as power method. The analytical solution for different types of regions that compose the reactor is obtained, such as fuel and reflector regions. Four average fluxes in the node surfaces are used as boundary conditions for analytical solution. Discontinuity factors on the node surfaces derived from the homogenization process are applied to maintain averages reaction rates and the net current in the fuel assembly (FA). To validate the results obtained by the analytical solution a relative power density distribution in the FAs is determined from the neutron flux distribution and compared with the reference values. The results show good accuracy and efficiency.

Thermal neutron detection by means of an organic solid-state track detector

International Nuclear Information System (INIS)

Doerschel, B.; Streubel, G.

1979-01-01

Thermal neutrons can be detected by means of organic solid-state track detectors if they are combined with radiators in which charged secondary particles are produced in neutron interaction processes. The secondary particles can produce etchable tracks in the detector material. For thermal neutron fluence determination from the track densities, the thermal neutron sensitivity was calculated for cellulose triacetate detectors with LiF radiators, taking into account energy and angular distribution of the alpha particles produced in the LiF radiator. This value is in good agreement with the sensitivity measured during irradiation in different neutron fields if corrections are considered the production of etchable or visuable tracks. Measuring range and measuring accuracy meet the requirements of thermal neutron detection in personnel dosimetry. Possibilities of extending the measuring range are discussed. (author)

New thermal neutron calibration channel at LNMRI/IRD

Energy Technology Data Exchange (ETDEWEB)

Astuto, A.; Lopes, R.T., E-mail: achillesbr@gmail.com [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Patrao, K.C.S.; Fonseca, E.S.; Pereira, W.W. [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ/LNMRI), Rio de Janeiro, RJ (Brazil). Lab. Nacional de Metrologia das Radiacoes Ionizantes

2015-07-01

A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four {sup 241}Am-Be sources with 596 GBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence (less than 1%) in the central chamber. (author)

Thermal neutron actinide data

International Nuclear Information System (INIS)

Tellier, H.

1992-01-01

During the 70's, the physicists involved in the cross section measurements for the low energy neutrons were almost exclusively interested in the resonance energy range. The thermal range was considered as sufficiently known. In the beginning of the 80's, reactor physicists had again to deal with the delicate problem of the power reactor temperature coefficient, essentially for the light water reactors. The measured value of the reactivity temperature coefficient does not agree with the computed one. The later is too negative. For obvious safety reasons, it is an important problem which must be solved. Several causes were suggested to explain this discrepancy. Among all these causes, the spectral shift in the thermal energy range seems to be very important. Sensibility calculations shown that this spectral shift is very sensitive to the shape of the neutron cross sections of the actinides for energies below one electron-volt. Consequently, reactor physicists require new and accurate measurements in the thermal and subthermal energy ranges. A part of these new measurement results were recently released and reviewed. The purpose of this study is to complete the preceding review with the new informations which are now available. In reactor physics the major actinides are the fertile nuclei, uranium 238, thorium 232 and plutonium 240 and the fissile nuclei, uranium 233, uranium 235 and plutonium 239. For the fertile nuclei the main datum is the capture cross section, for the fissile nuclei the data of interest are nu-bar, the fission and capture cross sections or a combination of these data such as η or α. In the following sections, we will review the neutron data of the major actinides for the energy below 1 eV

One-dimensional reactor kinetics model for RETRAN

International Nuclear Information System (INIS)

Gose, G.C.; Peterson, C.E.; Ellis, N.L.; McClure, J.A.

1981-01-01

This paper describes a one-dimensional spatial neutron kinetics model that was developed for the RETRAN code. The RETRAN -01 code has a point kinetics model to describe the reactor core behavior during thermal-hydraulic transients. A one-dimensional neutronics model has been developed for RETRAN-02. The ability to account for flux shape changes will permit an improved representation of the thermal and hydraulic feedback effects for many operational transients. 19 refs

A low noise ASIC for two dimensional neutron gas detector with performance of high spatial resolution (Contract research)

International Nuclear Information System (INIS)

Yamagishi, Hideshi; Toh, Kentaro; Nakamura, Tatsuya; Sakasai, Kaoru; Soyama, Kazuhiko

2012-02-01

An ASD-ASIC (Amplifier-Shaper-Discriminator ASIC) with fast response and low noise performances has been designed for two-dimensional position sensitive neutron gas detectors (InSPaD). The InSPaD is a 2D neutron detector system with 3 He gas and provides a high spatial resolution by making distinction between proton and triton particles generated in the gas chamber. The new ASD-ASIC is required to have very low noise, a wide dynamic range, good output linearity and high counting rate. The new ASD-ASIC has been designed by using CMOS and consisted of 64-channel ASDs, a 16-channel multiplexer with LVTTL drivers and sum amplifier system for summing all analog signals. The performances were evaluated by the Spice simulation. It was confirmed that the new ASD-ASIC had very low noise performance, wide dynamic range and fast signal processing functions. (author)

Cell death following thermal neutron exposure

Energy Technology Data Exchange (ETDEWEB)

Paterson, L.C. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Atanackovic, J. [Ontario Power Generation, Toronto, Ontario (Canada); Boyer, C. [Canadian Neutron Beam Centre, Chalk River, Ontario (Canada); El-Jaby, S.; Priest, N.D. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Seymour, C.B.; Boreham, D.R. [McMaster Univ., Hamilton, Ontario (Canada); Richardson, R.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2014-07-01

When individuals are exposed to unknown external ionizing radiation, it is desirable to have the means to assess both the absorbed dose received (Gy) and the radiation quality. Yet, conventional biodosimetry techniques, specifically the dicentric chromosome assay, cannot differentiate between the damage caused by high- and low-linear energy transfer (LET) exposures. Frequencies of apoptosis and necrosis, may provide an alternative method that assesses both the absorbed dose and radiation quality after unknown exposures. For this preliminary study, human lymphocytes were irradiated with {sup 60}Co gamma rays and thermal neutrons. Both apoptosis and necrosis increased with increasing gamma dose. In contrast, no dose-response was observed following thermal neutron exposure at doses up to 2.61 Gy. (author)

Thermal ignition revisited with two-dimensional molecular dynamics: role of fluctuations in activated collisions

OpenAIRE

Sirmas, Nick; Radulescu, Matei I.

2016-01-01

The problem of thermal ignition in a homogeneous gas is revisited from a molecular dynamics perspective. A two-dimensional model is adopted, which assumes reactive disks of type A and B in a fixed area that react to form type C products if an activation threshold for impact is surpassed. Such a reaction liberates kinetic energy to the product particles, representative of the heat release. The results for the ignition delay are compared with those obtained from the continuum description assumi...

Effect of Rotation for Two-Temperature Generalized Thermoelasticity of Two-Dimensional under Thermal Shock Problem

Directory of Open Access Journals (Sweden)

Kh. Lotfy

2013-01-01

Full Text Available The theory of two-temperature generalized thermoelasticity based on the theory of Youssef is used to solve boundary value problems of two-dimensional half-space. The governing equations are solved using normal mode method under the purview of the Lord-Şhulman (LS and the classical dynamical coupled theory (CD. The general solution obtained is applied to a specific problem of a half-space subjected to one type of heating, the thermal shock type. We study the influence of rotation on the total deformation of thermoelastic half-space and the interaction with each other under the influence of two temperature theory. The material is homogeneous isotropic elastic half-space. The methodology applied here is use of the normal mode analysis techniques that are used to solve the resulting nondimensional coupled field equations for the two theories. Numerical results for the displacement components, force stresses, and temperature distribution are presented graphically and discussed. The conductive temperature, the dynamical temperature, the stress, and the strain distributions are shown graphically with some comparisons.

Thermal neutron capture cross sections of tellurium isotopes

International Nuclear Information System (INIS)

Tomandl, I.; Honzatko, J.; Egidy, T. von; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

2003-01-01

New values for thermal neutron capture cross sections of the tellurium isotopes 122 Te, 124 Te, 125 Te, 126 Te, 128 Te, and 130 Te are reported. These values are based on a combination of newly determined partial γ-ray cross sections obtained from experiments on targets contained natural Te and γ intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given

Thermal neutron capture cross sections of tellurium isotopes

International Nuclear Information System (INIS)

Tomandl, I.; Honzatko, J.; Egidy, T. von; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

2004-01-01

New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given

Thermal neutron capture cross sections of tellurium isotopes

Energy Technology Data Exchange (ETDEWEB)

Tomandl, I.; Honzatko, J.; von Egidy, T.; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

2004-03-01

New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given.

DUNBID, the Delft University neutron backscattering imaging detector

International Nuclear Information System (INIS)

Bom, V.R.; Eijk, C.W.E. van; Ali, M.A.

2005-01-01

In the search for low-metallic land mines, the neutron backscattering technique may be applied if the soil is sufficiently dry. An advantage of this method is the speed of detection: the scanning speed may be made comparable to that of a metal detector. A two-dimensional position sensitive detector is tested to obtain an image of the back scattered thermal neutron radiation. Results of experiments using a radionuclide neutron source are presented. The on-mine to no-mine signal ratio can be improved by the application of a window on the neutron time-of-flight. Results using a pulsed neutron generator are also presented

Critical Behaviour of a Two-Dimensional Random Antiferromagnet

DEFF Research Database (Denmark)

Als-Nielsen, Jens Aage; Birgeneau, R. J.; Guggenheim, H. J.

1976-01-01

A neutron scattering study of the order parameter, correlation length and staggered susceptibility of the two-dimensional random antiferromagnet Rb2Mn0.5Ni0.5F4 is reported. The system is found to exhibit a well-defined phase transition with critical exponents identical to those of the isomorphou...... pure materials K2NiF4 and K2MnF4. Thus, in these systems, which have the asymptotic critical behaviour of the two-dimensional Ising model, randomness has no measurable effect on the phase-transition behaviour....

Impact of thermal and intermediate energy neutrons on the semiconductor memories for the CERN accelerators

CERN Document Server

Cecchetto, Matteo; Gerardin, Simone

A wide quantity of SRAM memories are employed along the Large Hadron Collider (LHC), the main CERN accelerator, and they are subjected to high levels of ionizing radiations which compromise the reliability of these devices. The Single Event Effect (SEE) qualification for components to be used in the complex high-energy accelerator at CERN relies on the characterization of two cross sections: 200-MeV protons and thermal neutrons. However, due to cost and time constraints, it is not always possible to characterize the SEE response of components to thermal neutrons, which is often regarded as negligible for components without borophosphosilicate glass (BPSG). Nevertheless, as recent studies show, the sensitivity of deep sub-micron technologies to thermal neutrons has increased owing to the presence of Boron 10 as a dopant and contact contaminant. The very large thermal neutron fluxes relative to high-energy hadron fluxes in some of the heavily shielded accelerator areas imply that even comparatively small therm...

Approximate solutions for the two-dimensional integral transport equation. Solution of complex two-dimensional transport problems

International Nuclear Information System (INIS)

Sanchez, Richard.

1980-11-01

This work is divided into two parts: the first part deals with the solution of complex two-dimensional transport problems, the second one (note CEA-N-2166) treats the critically mixed methods of resolution. A set of approximate solutions for the isotropic two-dimensional neutron transport problem has been developed using the interface current formalism. The method has been applied to regular lattices of rectangular cells containing a fuel pin, cladding, and water, or homogenized structural material. The cells are divided into zones that are homogeneous. A zone-wise flux expansion is used to formulate a direct collision probability problem within a cell. The coupling of the cells is effected by making extra assumptions on the currents entering and leaving the interfaces. Two codes have been written: CALLIOPE uses a cylindrical cell model and one or three terms for the flux expansion, and NAUSICAA uses a two-dimensional flux representation and does a truly two-dimensional calculation inside each cell. In both codes, one or three terms can be used to make a space-independent expansion of the angular fluxes entering and leaving each side of the cell. The accuracies and computing times achieved with the different approximations are illustrated by numerical studies on two benchmark problems and by calculations performed in the APOLLO multigroup code [fr

Two-dimensional simulation of the thermal stress effect on static and dynamic VDMOS characteristics

International Nuclear Information System (INIS)

Alwan, M.; Beydoun, B.; Ketata, K.; Zoaeter, M.

2005-01-01

Using a two-dimensional simulator, the effect of the thermal stress on static and dynamic vertical double-diffusion metal oxide semiconductor (VDMOS) characteristics have been investigated. The use of the device under certain thermal stress conditions can produce modifications of its physical and electrical properties. Based on physics and 2D simulations, this paper proposes an analysis of this stress effect observed on the electrical characteristics of the device. Parameters responsible of these modifications are determined. Approximate expressions of the ionization coefficients and breakdown voltage in terms of temperature are proposed. Non-punch-through junction theory is used to express the breakdown voltage and the space charge extension with respect to the impurity concentration and the temperature. The capacitances of the device have been also studied. The effect of the stress on C-V characteristics is observed and analyzed. We notice that the drain-gate, drain-source and gate-source capacitances are shifted due to the degradation of device physical properties versus thermal stress

Neutron scattering investigation on low-dimensional, quantum and frustrated magnetism and utilization of neutron polarization analysis. My first encounter with neutron research

International Nuclear Information System (INIS)

Kakurai, Kazuhisa

2013-01-01

My first encounter with neutron scattering research on low-dimensional magnetism at the Hahn-Meitner Institut under the supervision of Prof. H. Dachs and Prof. M. Steiner, were it all began, is accounted for. The polarized neutron analysis research on low-dimensional magnetism at the Institut Laue Langevin under the supervision of Dr. R. Pynn is also reported. I would like to dedicate this article to late Prof. H. Dachs expressing may deepest gratitude for his warm guidance during the early period of my neutron science carrier. (author)

Magnetic structure of two- and three-dimensional supramolecular compounds

Energy Technology Data Exchange (ETDEWEB)

Decurtins, S.; Schmalle, H.W.; Pellaux, R. [Zurich Univ. (Switzerland); Fischer, P.; Fauth, F. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Ouladdiaf, B. [Institut Max von Laue - Paul Langevin, 75 - Paris (France)

1997-09-01

Supramolecular chiral networks of oxalato-bridged transition metals show either two- or three-dimensional structural features. The magnetic structures of such compounds have been investigated by means of elastic neutron powder diffraction. (author) 2 figs., 2 refs.

Performance of a thermal neutron radiographic system using imaging plates

International Nuclear Information System (INIS)

Silvani, Maria Ines; Almeida, Gevaldo L. de; Furieri, Rosanne; Lopes, Ricardo T.

2009-01-01

A performance evaluation of a neutron radiographic system equipped with a thermal neutron sensitive imaging plate has been undertaken. It includes the assessment of spatial resolution, linearity, dynamic range and the response to exposure time, as well as a comparison of these parameters with the equivalent ones for neutron radiography employing conventional films and a gadolinium foil as converter. The evaluation and comparison between the radiographic systems have been performed at the Instituto de Engenharia Nuclear - CNEN, using the Argonauta Reactor as source of thermal neutrons and a commercially available imaging plate reader. (author)

Enhancement of thermal neutron attenuation of nano-B4C, -BN dispersed neutron shielding polymer nanocomposites

International Nuclear Information System (INIS)

Kim, Jaewoo; Lee, Byung-Chul; Uhm, Young Rang; Miller, William H.

2014-01-01

Highlights: • Preparation of B 4 C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B 4 C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B 4 C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.H.

1993-01-01

In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-A19897, R.H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280degF. Table 1 lists the neutron shield materials tested. The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found. The Bisco modified NS-4 and Reactor Experiments HMPP are both acceptable materials from a thermal accident standpoint for use in the shipping cask. Tests of the Kobe PP-R01 and Envirotech HDPE were stopped for safety reasons, due to inability to deal with the heavy smoke, before completion of the 30-minute heating phase. However these materials may prove satisfactory if they could undergo the complete heating. (J.P.N.)

Neutron thermalization in light water

International Nuclear Information System (INIS)

Abbate, M.J.; Lolich, J.V.

1975-05-01

Investigations related to neutron thermalization in light water have been made. Neutron spectra under quasi-infinite-medium conditions have been measured by the time-of-flight technique and calculations were performed with different codes. Through the use of improved experimental techniques and the best known calculational techniques available, the known discrepancies between experimentals and theoretical values were below from 40% to 16%. The present disagreement is believed to be due the scattering model used (ENDF-GASKET, based on the modified Haywood II frequency spectra), that shows to be very satisfactory for poisoned light water cases. Moreover, previous experiments were completed and differential, integral and pulse-source experimental techniques were improved. Also a second step of a neutron and reactor calculation system was completed. (author)

Introduction to the theory of thermal neutron scattering

CERN Document Server

Squires, G L

2012-01-01

Since the advent of the nuclear reactor, thermal neutron scattering has proved a valuable tool for studying many properties of solids and liquids, and research workers are active in the field at reactor centres and universities throughout the world. This classic text provides the basic quantum theory of thermal neutron scattering and applies the concepts to scattering by crystals, liquids and magnetic systems. Other topics discussed are the relation of the scattering to correlation functions in the scattering system, the dynamical theory of scattering and polarisation analysis. No previous knowledge of the theory of thermal neutron scattering is assumed, but basic knowledge of quantum mechanics and solid state physics is required. The book is intended for experimenters rather than theoreticians, and the discussion is kept as informal as possible. A number of examples, with worked solutions, are included as an aid to the understanding of the text.

ATLAS-TPX: a two-layer pixel detector setup for neutron detection and radiation field characterization

International Nuclear Information System (INIS)

Bergmann, B.; Caicedo, I.; Pospisil, S.; Vykydal, Z.; Leroy, C.

2016-01-01

A two-layer pixel detector setup (ATLAS-TPX), designed for thermal and fast neutron detection and radiation field characterization is presented. It consists of two segmented silicon detectors (256 × 256 pixels, pixel pitch 55 μm, thicknesses 300 μm and 500 μm) facing each other. To enhance the neutron detection efficiency a set of converter layers is inserted in between these detectors. The pixelation and the two-layer design allow a discrimination of neutrons against γs by pattern recognition and against charged particles by using the coincidence and anticoincidence information. The neutron conversion and detection efficiencies are measured in a thermal neutron field and fast neutron fields with energies up to 600 MeV. A Geant4 simulation model is presented, which is validated against the measured detector responses. The reliability of the coincidence and anticoincidence technique is demonstrated and possible applications of the detector setup are briefly outlined.

Investigation of the response of a neutron moisture meter using a multigroup, two-dimensional diffusion theory code

International Nuclear Information System (INIS)

Ritchie, A.I.M.; Wilson, D.J.

1984-12-01

A multigroup diffusion code has been used to predict the count rate from a neutron moisture meter for a range of values of soil water content ω, thermal neutron absorption cross section Ssub(a) (defined as Σsub(a)/rho) of the soil matrix and soil matrix density rho. Two dimensions adequately approximated the geometry of the source, detector and soil surrounding the detector. Seven energy groups, the data for which were condensed from 128 group data set over the neutron energy spectrum appropriate to the soil-water mixture under study, proved adequate to describe neutron slowing-down and diffusion. The soil-water mixture was an SiO 2 →water mixture, with the absorption cross section of SiO 2 increased to cover the range of Σsub(a) required. The response to changes in matrix density is, in general, linear but the response to changes in water content is not linear over the range of parameter values investigated. Tabular results are presented which allow interpolation of the response for a particular ω, Ssub(a) and rho. It is shown that R(ω, Ssub(a), rho) rho M(Ssub(a)) + C(ω) is a crude representation of the response over a very limited range of variation of ω, and Ssub(a). As the response is a slowly varying function of rho, Ssub(a) and ω, a polynomial fit will provide a better estimate of the response for values of rho, Ssub(a) and ω not tabulated

System and plastic scintillator for discrimination of thermal neutron, fast neutron, and gamma radiation

Science.gov (United States)

Zaitseva, Natalia P.; Carman, M. Leslie; Faust, Michelle A.; Glenn, Andrew M.; Martinez, H. Paul; Pawelczak, Iwona A.; Payne, Stephen A.

2017-05-16

A scintillator material according to one embodiment includes a polymer matrix; a primary dye in the polymer matrix, the primary dye being a fluorescent dye, the primary dye being present in an amount of 3 wt % or more; and at least one component in the polymer matrix, the component being selected from a group consisting of B, Li, Gd, a B-containing compound, a Li-containing compound and a Gd-containing compound, wherein the scintillator material exhibits an optical response signature for thermal neutrons that is different than an optical response signature for fast neutrons and gamma rays. A system according to one embodiment includes a scintillator material as disclosed herein and a photodetector for detecting the response of the material to fast neutron, thermal neutron and gamma ray irradiation.

Thermal structure of the ionosphere of Mars - simulations with one- and two-dimensional models

International Nuclear Information System (INIS)

Singhal, R.P.; Whitten, R.C.

1988-01-01

Heat flux saturation effects are included in the present one- and two-dimensional models of the Martian upper ionosphere's thermal structure. The inclusion of small upper boundary and volume heat sources is found to yield satisfactory simulations of the dayside ion temperature observation results obtained by Viking 1's retarding potential analyzers. It is noted that the plasma flow-transport of heat from the dayside to the nightside makes no contribution to the ion and electron temperatures that have been calculated for the nightside. 22 references

Filtered thermal neutron captured cross sections measurements and decay heat calculations

International Nuclear Information System (INIS)

Pham Ngoc Son; Vuong Huu Tan

2015-01-01

Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R ed ) of 420 and neutron flux (Φ th ) of 1.6*10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross sections for nuclide of 51 V, by the activation method relative to the standard reaction 197 Au(n,γ) 198 Au. In addition to the activities of neutron capture cross sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U are introduced in this report. (author)

Preparation of rock samples for measurement of the thermal neutron macroscopic absorption cross-section

International Nuclear Information System (INIS)

Czubek, J.A.; Burda, J.; Drozdowicz, K.; Igielski, A.; Kowalik, W.; Krynicka-Drozdowicz, E.; Woznicka, U.

1986-03-01

Preparation of rock samples for the measurement of the thermal neutron macroscopic absorption cross-section in small cylindrical two-region systems by a pulsed technique is presented. Requirements which should be fulfilled during the preparation of the samples due to physical assumptions of the method are given. A cylindrical vessel is filled with crushed rock and saturated with a medium strongly absorbing thermal neutrons. Water solutions of boric acid of well-known macroscopic absorption cross-section are used. Mass contributions of the components in the sample are specified. This is necessary for the calculation of the thermal neutron macroscopic absorption cross-section of the rock matrix. The conditions necessary for assuring the required accuracy of the measurement are given and the detailed procedure of preparation of the rock sample is described. (author)

Need for nuclear data for thermal neutron reactors

International Nuclear Information System (INIS)

Bouchard, J.; Golinelli, C.; Tellier, H.

1983-01-01

The need for nuclear data for thermal neutron reactors is conditioned by the persisting lack of agreement between the calculation and measurement of certain parameters, by the benefit that can be drawn from reduction of the marginal areas and by envisaged modifications. Three particular fields are delineated. Reduction of the deviation in temperature coefficients by modification of the shape of the effective capture cross sections of uranium-238 and -235 in the thermal range. The increase in precision of kinetic measurements by a better knowledge of data connected to slowed-down neutrons. Improvement in predicting the neutron activity of the fuels used in measuring the effective capture cross sections of plutonium-242 and americium-243. (Auth.)

On the theory of ultracold neutrons scattering by Davydov solitons

International Nuclear Information System (INIS)

Brizhik, L.S.

1984-01-01

Elastic coherent scattering of ultracold neutrons by Davydov solitons in one-dimensional periodic molecular chains without account of thermal oscillations of chain atoms is studied. It is shown that the expression for the differential cross section of the elastic neutron scattering by Davydov soliton breaks down into two components. One of them corresponds to scattering by a resting soliton, the other is proportional to the soliton velocity and has a sharp maximum in the direction of mirror reflection of neutrons from the chain

Measurement of thermal neutron spectra using LINAC in Japan Atomic Energy Research Institute (JAERI)

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1982-01-01

The exact grasp of thermal neutron spectra in a core region is very important for obtaining accurate thermal neutron group constants in the calculation for the nuclear design of a reactor core. For the accurate grasp of thermal neutron spectra, the capability of thermal neutron spectra to describe the moderator cross-sections for thermal neutron scattering is a key factor. Accordingly, 0 deg angular thermal neutron spectra were measured by the time of flight (TOF) method using the JAERI LINAC as a pulsed neutron source, for light water system added with Cd and In, high temperature graphite system added with boron, and light water-natural uranium heterogeneous multiplication system among the reactor moderators of light water or graphite systems. First, the equations to give the time of flight and neutron flux by TOF method were analyzed, and several corrections were investigated, such as those for detector efficiency, background, the transmission coefficient of air and the Al window of a flight tube, mean emission time of neutrons, and the distortion effect of re-entrant hole on thermal neutron spectra. Then, the experimental system, results and calculation were reported for the experiments on the above three moderator systems. Finally, the measurement of fast neutron spectra in natural uranium system and that of the efficiency of a 6 Li glass scintillator detector are described. (Wakatsuki, Y.)

Two-dimensional position sensitive neutron detector

Indian Academy of Sciences (India)

The detector is a 3He + Kr filled multiwire proportional counter with charge division position readout and has a sensitive area of 345 mm × 345 mm, pixel size 5 mm × 5 mm, active depth 25 mm and is designed for efficiency of 70% for 4 Å neutrons. The detector is tested with 0.5 bar 3He + 1.5 bar krypton gas mixture in active ...

Assessment of the impact of neutronic/thermal-hydraulic coupling on the design and performance of nuclear reactors for space propulsion

International Nuclear Information System (INIS)

Aithal, S.M.; Aldemir, T.; Vafai, K.

1994-01-01

A series of studies has been performed to investigate the potential impact of the coupling between neutronics and thermal hydraulics on the design and performance assessment of solid core reactors for nuclear thermal space propulsion, using the particle bed reactor (PBR) concept as an example system. For a given temperature distribution in the reactor, the k eff and steady-state core power distribution are obtained from three-dimensional, continuous energy Monte Carlo simulations using the MCNP code. For a given core power distribution, determination of the temperature distribution in the core and hydrogen-filled annulus between the reflector and pressure vessel is based on a nonthermal equilibrium analysis. The results show that a realistic estimation of fuel, core size, and control requirements for PBRs using hydrogenous moderators, as well as optimization of the overall engine design, may require coupled neutronic/thermal-hydraulic studies. However, it may be possible to estimate the thermal safety margins and propellant exit temperatures based on power distributions obtained from neutronic calculations at room temperature. The results also show that, while variation of the hydrogen flow rate in the annulus has been proposed as a partial control mechanism for PBRs, such control mechanism may not be feasible for PBRs with high moderator-to-fuel ratios and hence soft core neutron spectra

VHTR core modeling: coupling between neutronic and thermal-hydraulics

International Nuclear Information System (INIS)

Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

2005-01-01

Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

Thermal neutron group constants in monoatomic-gas approximation

Energy Technology Data Exchange (ETDEWEB)

Matausek, M V; Bosevski, T [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-12-15

To solve the problem of space-energy neutron distribution in an elementary reactor cell, a combination of the multigroup procedure and the P{sub 3} approximation of the spherical harmonics method was chosen. The calculation was divided into two independent parts: the first part was to provide multigroup constants which serve as input data for the second part - the determination of the slow neutron spectra. In the present report only the first part of the problem will be discussed. The velocity dependence of cross-sections and scattering function in thermal range was interpreted by the monoatomic-gas model. A digital computer program was developed for the evaluation of the group values for these quantities (author00.

Indirect and direct measurement of thermal neutron acceleration by inelastic scattering on the 177Lu isomer

International Nuclear Information System (INIS)

Belier, G.; Roig, O.; Meot, V.; Daugas, J.M.; Aupiais, J.; Jutier, Ch.; Le Petit, G.; Veyssiere, Ch.

2008-01-01

When neutrons interact with isomers, these isomers can de-excite and in such a reaction the outgoing neutron has an energy greater than the in-going one. This process is referred as Inelastic Neutron Acceleration or Super-elastic Scattering. Up to now this process was observed for only two nucleus, 152m Eu and 180m Hf by measuring the number of fast neutrons produced by isomeric targets irradiated with thermal neutrons. In these experiments the energies of the accelerated neutrons were not measured. This report presents an indirect measurement of inelastic neutron acceleration on 177m Lu, based on the burn-up and the radiative capture cross sections measurements. Since at thermal energies the inelastic scattering and the radiative capture are the only processes that contribute to the isomer burn-up, the inelastic cross section can be deduced from the difference between the two measured quantities. Applying this method for the 177 Lu isomer with different neutron fluxes we obtained a value of (257 ± 50) barns (for a temperature of 323 K) and determined that there is no integral resonance for this process. In addition the radiative capture cross section on 177g Lu was measured with a much better accuracy than the accepted value. Since the acceleration cross section is quite high, a direct measurement of this process was undertaken, sending thermal neutrons and measuring the fast neutrons. The main goal now is to measure the outgoing neutron energies in order to identify the neutron transitions in the exit channel. In particular the K conservation question can be addressed by such a measurement. (author)

Targets for bulk hydrogen analysis using thermal neutrons

CERN Document Server

Csikai, J; Buczko, C M

2002-01-01

The reflection property of substances can be characterized by the reflection cross-section of thermal neutrons, sigma subbeta. A combination of the targets with thin polyethylene foils allowed an estimation of the flux depression of thermal neutrons caused by a bulk sample containing highly absorbing elements or compounds. Some new and more accurate sigma subbeta values were determined by using the combined target arrangement. For the ratio, R of the reflection and the elastic scattering cross-sections of thermal neutrons, R=sigma subbeta/sigma sub E sub L a value of 0.60+-0.02 was found on the basis of the data obtained for a number of elements from H to Pb. Using this correlation factor, and the sigma sub E sub L values, the unknown sigma subbeta data can be deduced. The equivalent thicknesses, to polyethylene or hydrogen, of the different target materials were determined from the sigma subbeta values.

KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

International Nuclear Information System (INIS)

Shamasundar, B.I.; Fehrenbach, M.E.

1981-05-01

The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

Neutron radiography imaging with 2-dimensional photon counting method and its problems

International Nuclear Information System (INIS)

Ikeda, Y.; Kobayashi, H.; Niwa, T.; Kataoka, T.

1988-01-01

A ultra sensitive neutron imaging system has been deviced with a 2-dimensional photon counting camara (ARGUS 100). The imaging system is composed by a 2-dimensional single photon counting tube and a low background vidicon followed with an image processing unit and frame memories. By using the imaging system, electronic neutron radiography (NTV) has been possible under the neutron flux less than 3 x 10 4 n/cm 2 ·s. (author)

Methods and apparatus for environmental correction of thermal neutron logs

International Nuclear Information System (INIS)

Preeg, W.E.; Scott, H.D.

1983-01-01

An on-line environmentally-corrected measurement of the thermal neutron decay time (tau) of an earth formation traversed by a borehole is provided in a two-detector, pulsed neutron logging tool, by measuring tau at each detector and combining the two tau measurements in accordance with a previously established empirical relationship of the general form: tau = tausub(F) +A(tausub(F) + tausub(N)B) + C, where tausub(F) and tausub(N) are the tau measurements at the far-spaced and near-spaced detectors, respectively, A is a correction coefficient for borehole capture cross section effects, B is a correction coefficient for neutron diffusion effects, and C is a constant related to parameters of the logging tool. Preferred numerical values of A, B and C are disclosed, and a relationship for more accurately approximating the A term to specific borehole conditions. (author)

Characterization of the internal background for thermal and fast neutron detection with CLLB

Energy Technology Data Exchange (ETDEWEB)

Woolf, Richard S., E-mail: richard.woolf@nrl.navy.mil; Phlips, Bernard F.; Wulf, Eric A.

2016-12-01

We report on a set of experiments conducted to determine what effects, if any, the internal background in the CLLB scintillation detector has on the thermal neutron detection performance. We conducted source measurements using an unmoderated and moderated {sup 252}Cf neutron/γ-ray source and long (48-h), unshielded and shielded, background measurements to characterize the internal background with and without a source present. These measurements allowed us to determine the 2-d event selections needed to isolate the thermal neutron peak observed in pulse shape vs. energy space and apply those selections to our background measurements. Our results indicate that the thermal neutron detection capabilities of the CLLB are marginally affected by the presence of internal background. An unmoderated 113-µCi {sup 252}Cf source at 15 cm from the detector yields a thermal neutron rate of 8×10{sup −2}/s cm{sup 3}, while moderating the source with 5 cm of polyethylene yields a thermal neutron rate of 5.5×10{sup −1}/s cm{sup 3}. The measured background rate for events that fall within the selected thermal neutron region is 1.2×10{sup −3}/s cm{sup 3}. Lastly, the potential for CLLB for detecting fast neutrons was investigated.

DWARF, 1-D Few-Group Neutron Diffusion with Thermal Feedback for Burnup and Xe Oscillation

International Nuclear Information System (INIS)

Anderson, E.C.; Putnam, G.E.

1975-01-01

1 - Description of problem or function: DWARF allows one-dimensional simulation of reactor burnup and xenon oscillation problems in slab, cylindrical, or spherical geometry using a few-group diffusion theory model. 2 - Method of solution: The few-group, neutron diffusion theory equations are reduced to a system of finite-difference equations that are solved for each group by the Gauss method at each time point. Fission neutron source iteration can be accelerated with Chebyshev extrapolation. A thermal feedback iterative loop is used to obtain consistent solutions for the distributions of reactor power, neutron flux, and fuel and coolant properties with the neutron group constants functions of the latter. Solutions for the new nuclide concentrations of a time-point are made with the flux assumed constant in the time interval. 3 - Restrictions on the complexity of the problem - Maxima of: 4 groups; 40 regions; 50 macroscopic materials (Only 10 are functions of the feedback variables); 50 nuclides per region; 250 mesh points

Measurement and simulation of thermal neutron flux distribution in the RTP core

Science.gov (United States)

Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

2018-01-01

The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

Thermal-hydraulic and neutronic considerations for designing a lithium-cooled tokamak blanket

International Nuclear Information System (INIS)

Chao, J.; Mikic, B.; Todreas, N.

1978-12-01

A methodology for the design of lithium cooled blankets is developed. The thermal-hydraulics, neutronics and interactions between them are extensively investigated. In thermal hydraulics, two models illustrate the methodology used to obtain the acceptable ranges for a set of design parameters. The methodology can be used to identify the limiting constraints for a particular design. A complete neutronic scheme is set up for the calculations of the volumetric heating rate as a function of the distance from the first wall, the breeding ratio as a function of the amount of structural material in the blanket, and the radiation damage in terms of atom displacements and gas production rate. Different values of the volume percent of Type-316 stainless steel are assigned in four breeding zones to represent a nonuniformly distributed structural material which satisfies various thermal-hydraulic requirements. The role that the radiation damage plays in the overall design methodology is described. The product of the first wall lifetime and neutron loading is limited by the radiation damage which degrades the mechanical properties of the material

Fast and thermal neutron intensity measurements at the KFUPM PGNAA setup

CERN Document Server

Al-Jarallah, M I; Fazal-Ur-Rehman; Abu-Jarad, F A

2002-01-01

Fast and thermal neutron intensity distributions have been measured at an accelerator based prompt gamma ray neutron activation analysis (PGNAA) setup. The setup is built at the 350 keV accelerator laboratory of King Fahd University of Petroleum and Minerals (KFUPM). The setup is mainly designed to carry out PGNAA elemental analysis via thermal neutron capture. In this study relative intensity of fast and thermal neutrons was measured as a function of the PGNAA moderator assembly parameters using nuclear track detectors (NTDs). The relative intensity of the neutrons was measured inside the sample region as a function of front moderator thickness as well as sample length. Measurements were carried out at the KFUPM 350 keV accelerator using 2.8 MeV pulsed neutron beam from D(d,n) reaction. The pulsed deuteron beam with 5 ns pulse width and 30 kHz frequency was used to produce neutrons. Experimental results were compared with results of Monte Carlo design calculations of the PGNAA setup. A good agreement has bee...

Fast and thermal neutron intensity measurements at the KFUPM PGNAA setup

Energy Technology Data Exchange (ETDEWEB)

Al-Jarallah, M.I.; Naqvi, A.A. E-mail: aanaqvi@kfupm.edu.sa; Fazal-ur-Rehman; Abu-jarad, F

2002-10-01

Fast and thermal neutron intensity distributions have been measured at an accelerator based prompt gamma ray neutron activation analysis (PGNAA) setup. The setup is built at the 350 keV accelerator laboratory of King Fahd University of Petroleum and Minerals (KFUPM). The setup is mainly designed to carry out PGNAA elemental analysis via thermal neutron capture. In this study relative intensity of fast and thermal neutrons was measured as a function of the PGNAA moderator assembly parameters using nuclear track detectors (NTDs). The relative intensity of the neutrons was measured inside the sample region as a function of front moderator thickness as well as sample length. Measurements were carried out at the KFUPM 350 keV accelerator using 2.8 MeV pulsed neutron beam from D(d,n) reaction. The pulsed deuteron beam with 5 ns pulse width and 30 kHz frequency was used to produce neutrons. Experimental results were compared with results of Monte Carlo design calculations of the PGNAA setup. A good agreement has been found between the experimental results and the calculations.

Filtered thermal neutron captured cross-sections measurements and decay heat calculations

International Nuclear Information System (INIS)

Son, Pham Ngoc; Tan, Vuong Huu

2014-01-01

Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R cd ) of 420 and neutron flux (Φ th ) of 1.6x10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross-sections for nuclide of 51 V, 55 Mn, 180 Hf and 186 W by the activation method relative to the standard reaction 197 Au(n,g) 198 Au. In addition to the activities of neutron capture cross-sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U, 238 U, 239 Pu and 232 Th are introduced in this report. (author)

Enhancement of thermal neutron attenuation of nano-B{sub 4}C, -BN dispersed neutron shielding polymer nanocomposites

Energy Technology Data Exchange (ETDEWEB)

Kim, Jaewoo, E-mail: kimj@kaeri.re.kr [Nuclear Materials Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); WCI Quantum Beam based Radiation Research Center, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States); Lee, Byung-Chul [Nuclear Reactor Core Design Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Uhm, Young Rang [Radioisotopes Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Miller, William H. [Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States)

2014-10-15

Highlights: • Preparation of B{sub 4}C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B{sub 4}C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B{sub 4}C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption.

Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

International Nuclear Information System (INIS)

Duran, I.; Bolshakova, I.; Holyaka, R.; Viererbl, L.; Lahodova, Z.; Sentkerestiova, J.; Bem, P.

2010-01-01

We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10 16 cm -2 was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

Investigation of coupling scheme for neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.

1988-01-01

Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described

High temperature ductility of austenitic alloys exposed to thermal neutrons

International Nuclear Information System (INIS)

Watanabe, K.; Kondo, T.; Ogawa, Y.

1982-01-01

Loss of high temperature ductility due to thermal neutron irradiation was examined by slow strain rate test in vacuum up to 1000 0 C. The results on two heats of Hastelloy alloy X with different boron contents were analyzed with respect to the influence of the temperatures of irradiation and tensile tests, neutron fluence and the associated helium production due to nuclear transmutation reaction. The loss of ductility was enhanced by increasing either temperature or neutron fluence. Simple extrapolations yielded the estimated threshold fluence and the end-of-life ductility values at 900 and 1000 0 C in case where the materials were used in near-core regions of VHTR. The observed relationship between Ni content and the ductility loss has suggested a potential utilization of Fe-based alloys for seathing of the neutron absorber materials

Effect of Different Thermal Neutron Fluxes on Blood of Male Mice

International Nuclear Information System (INIS)

Abd El-Latif, A.A.; Saeid, Kh. S.; Abd El-Latif, A.A.; Emara, N.M.; Emara, N.M.

2010-01-01

This work deals with the exposing of male mice to different fluxes of thermal neutron .Investigation has been performed by calculating of thermal neutron fluxes(0.27x10 8 N/cm 2 . 1h , 0.54x10 8 N/cm 2 . 1h, 1.08x10 8 N/cm 2 . 1h, 2.16x10 8 N/cm 2 . 3h and 4.32x10 8 N/cm 2 . 6h) which emitted from neutron irradiation cell with source Ra - Be (α,n) have activity 3 m. Ci made by leybold(55930) . The number and differential leucocytes counts types of white blood cells in million per cubic millimeter (W. B. Cs. mm -3 ) ,the number of platelets mm -3 ,the number of red blood cells in million per cubic millimeter (R. B. Cs. mm -3 ), the hemoglobin in Blood (mg/dl), the lymphocytes ,and the eosiniphil leucocytes in blood decrease with increasing thermal neutron fluxes. But neutrophile and monocytes in blood increase with increasing the thermal neutron fluxes

Verification of thermal-irradiation stress analytical code VIENUS of graphite block

International Nuclear Information System (INIS)

Iyoku, Tatsuo; Ishihara, Masahiro; Shiozawa, Shusaku; Shirai, Hiroshi; Minato, Kazuo.

1992-02-01

The core graphite components of the High Temperature Engineering Test Reactor (HTTR) show both the dimensional change (irradiation shrinkage) and creep behavior due to fast neutron irradiation under the temperature and the fast neutron irradiation conditions of the HTTR. Therefore, thermal/irradiation stress analytical code, VIENUS, which treats these graphite irradiation behavior, is to be employed in order to design the core components such as fuel block etc. of the HTTR. The VIENUS is a two dimensional finite element viscoelastic stress analytical code to take account of changes in mechanical properties, thermal strain, irradiation-induced dimensional change and creep in the fast neutron irradiation environment. Verification analyses were carried out in order to prove the validity of this code based on the irradiation tests of the 8th OGL-1 fuel assembly and the fuel element of the Peach Bottom reactor. This report describes the outline of the VIENUS code and its verification analyses. (author)

Analytic scattering kernels for neutron thermalization studies

International Nuclear Information System (INIS)

Sears, V.F.

1990-01-01

Current plans call for the inclusion of a liquid hydrogen or deuterium cold source in the NRU replacement vessel. This report is part of an ongoing study of neutron thermalization in such a cold source. Here, we develop a simple analytical model for the scattering kernel of monatomic and diatomic liquids. We also present the results of extensive numerical calculations based on this model for liquid hydrogen, liquid deuterium, and mixtures of the two. These calculations demonstrate the dependence of the scattering kernel on the incident and scattered-neutron energies, the behavior near rotational thresholds, the dependence on the centre-of-mass pair correlations, the dependence on the ortho concentration, and the dependence on the deuterium concentration in H 2 /D 2 mixtures. The total scattering cross sections are also calculated and compared with available experimental results

Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

International Nuclear Information System (INIS)

Chenu, A.

2011-10-01

Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

Comparison of the thermal neutron scattering treatment in MCNP6 and GEANT4 codes

Science.gov (United States)

Tran, H. N.; Marchix, A.; Letourneau, A.; Darpentigny, J.; Menelle, A.; Ott, F.; Schwindling, J.; Chauvin, N.

2018-06-01

To ensure the reliability of simulation tools, verification and comparison should be made regularly. This paper describes the work performed in order to compare the neutron transport treatment in MCNP6.1 and GEANT4-10.3 in the thermal energy range. This work focuses on the thermal neutron scattering processes for several potential materials which would be involved in the neutron source designs of Compact Accelerator-based Neutrons Sources (CANS), such as beryllium metal, beryllium oxide, polyethylene, graphite, para-hydrogen, light water, heavy water, aluminium and iron. Both thermal scattering law and free gas model, coming from the evaluated data library ENDF/B-VII, were considered. It was observed that the GEANT4.10.03-patch2 version was not able to account properly the coherent elastic process occurring in crystal lattice. This bug is treated in this work and it should be included in the next release of the code. Cross section sampling and integral tests have been performed for both simulation codes showing a fair agreement between the two codes for most of the materials except for iron and aluminium.

Simultaneous thermal neutron decay time and porosity logging system

International Nuclear Information System (INIS)

Schultz, W.E.; Smith, H.D.; Smith, M.P.

1980-01-01

An improved method and apparatus are described for simultaneously measuring the porosity and thermal neutron capture cross section of earth formations in situ in the vicinity of a well borehole using pulsed neutron well logging techniques. The logging tool which is moved through the borehole consists of a 14 MeV pulsed neutron source, an epithermal neutron detector and a combination gamma ray and fast neutron detector. The associated gating systems, counters and combined digital computer are sited above ground. (U.K.)

CaSO{sub 4}:Dy microphosphor for thermal neutron dosimetry

Energy Technology Data Exchange (ETDEWEB)

Bhadane, Mahesh S. [Microtron Accelerator Laboratory, Department of Physics, Savitribai Phule Pune University, Pune 411007 (India); Mandlik, Nandkumar [Department of Physics, Fergusson College, Savitribai Phule Pune University, Pune 411007 (India); Patil, B.J. [Department of Physics, Abasaheb Garware College, Pune 411004 (India); Dahiwale, S.S.; Sature, K.R.; Bhoraskar, V.N. [Microtron Accelerator Laboratory, Department of Physics, Savitribai Phule Pune University, Pune 411007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Microtron Accelerator Laboratory, Department of Physics, Savitribai Phule Pune University, Pune 411007 (India)

2016-02-15

Dysprosium-doped calcium sulphate (CaSO{sub 4}:Dy) microphosphor was synthesized by acid re-crystallization method and its thermoluminescence (TL) properties irradiated with thermal neutrons was studied. Structural and morphological characteristics have been studied using X-ray diffraction and SEM which mainly exhibits a orthorhombic structure with particle size of 200 to 250 µm. Moreover, thermal neutron dosimetric characteristics of the microphosphor such as thermoluminescence glow curve, TL dose–response have been studied. This microphosphor powder represents a TL glow peak (T{sub max}) centered at around 240 °C. The TL response of CaSO{sub 4}:Dy microphosphor as a function of thermal neutron fluence is observed to be very linear upto the fluence of 52×10{sup 11} n/cm{sup 2} and further saturates. In addition, TL glow curves were deconvoluted by computerized glow curve deconvolution (CGCD) method and corresponding trapping parameters have been determined. It has been found that for every deconvoluted peak there is change in the order of kinetics. Overall, the experimental results show that the CaSO{sub 4}:Dy microphosphor can have potential to be an effective thermal neutron dosimetry. - Highlights: • Acid-recrystallization method is used to prepare CaSO{sub 4}:Dy microphosphor • CaSO{sub 4}:Dy phosphor irradiated thermal neutrons for dosimetric application. • TL response curve showed to be a perfect linear. • Trapping parameters has been calculated using CGCD curve fitting.

Calculation of neutron flux distribution of thermal neutrons from microtron converter in a graphite moderator with water reflector

International Nuclear Information System (INIS)

Andrejsek, K.

1977-01-01

The calculation is made of the thermal neutron flux in the moderator and reflector by solving the neutron diffusion equation using the four-group theory. The correction for neutron absorption in the moderator was carried out using the perturbation theory. The calculation was carried out for four groups with the following energy ranges: the first group 2 MeV to 3 keV, the second group 3 keV to 5 eV, the third group 5 eV to 0.025 eV and the fourth group 0.025 eV. The values of the macroscopic cross section of capture and scattering, of the diffusion coefficient, the macroscopic cross section of the moderator, of the neutron age and the extrapolation length for the water-graphite moderator used in the calculations are given. The spatial distribution of the thermal neutron flux is graphically represented for graphite of a 30, 40, and 50 cm radius and for graphite of a 30 and 40 cm radius with a 10 cm water reflector; a graphic comparison is made of the distribution of the thermal neutron flux in water and in graphite, both 40 cm in radius. The system of graphite with reflector proved to be the best and most efficient system for raising the flux density of thermal neutrons. (J.P.)

Monte Carlo simulation of a coded-aperture thermal neutron camera

International Nuclear Information System (INIS)

Dioszegi, I.; Salwen, C.; Forman, L.

2011-01-01

We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm"2 active area "3He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in "3He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)

The measurement of thermal neutron constants of the soil; application to the calibration of neutron moisture gauges and to the pedological study of soil

International Nuclear Information System (INIS)

Couchat, P.; Marcesse, J.; Carre, C.; Le Ho, J.

1975-01-01

The neutronic method for measuring the water content of soils is more and more used by agronomists, hydrogeologists and pedologists. On the other hand the studies on the phenomena of slowing down and diffusion process have shown a narrow relation between the thermal absorption (Σ(a)) and diffusion (Σ(d)) constants and the thermal flux developed in the soil around a fast neutron source like Am-Be. Two original applications of the direct measurement of Σ(a) and Σ(d) are then presented. The method described consists in the measurement, in a cube of graphite with Am-Be source in the middle, on one side of the perturbation of the thermal flux, obtained by the introduction of 300g of soil, and on the other side of the transmitted thermal flux measured through the same sample of soil, on a side of the cube. After calibrating the device, these two parameters give Σ(a) and Σ(d) which are easily introduced in the calibration equation of neutron moisture gauge. Also these two values are useful for the pedologists because Σ(d) is connected to clay content in the soil and Σ(a) is connected to the type of clay by the way of rare earth contents [fr

Tailoring thermal conductivity via three-dimensional porous alumina.

Science.gov (United States)

Abad, Begoña; Maiz, Jon; Ruiz-Clavijo, Alejandra; Caballero-Calero, Olga; Martin-Gonzalez, Marisol

2016-12-09

Three-dimensional anodic alumina templates (3D-AAO) are an astonishing framework with open highly ordered three-dimensional skeleton structures. Since these templates are architecturally different from conventional solids or porous templates, they teem with opportunities for engineering thermal properties. By establishing the mechanisms of heat transfer in these frameworks, we aim to create materials with tailored thermal properties. The effective thermal conductivity of an empty 3D-AAO membrane was measured. As the effective medium theory was not valid to extract the skeletal thermal conductivity of 3D-AAO, a simple 3D thermal conduction model was developed, based on a mixed series and parallel thermal resistor circuit, giving a skeletal thermal conductivity value of approximately 1.25 W·m -1 ·K -1 , which matches the value of the ordinary AAO membranes prepared from the same acid solution. The effect of different filler materials as well as the variation of the number of transversal nanochannels and the length of the 3D-AAO membrane in the effective thermal conductivity of the composite was studied. Finally, the thermal conductivity of two 3D-AAO membranes filled with cobalt and bismuth telluride was also measured, which was in good agreement with the thermal model predictions. Therefore, this work proved this structure as a powerful approach to tailor thermal properties.

Tailoring thermal conductivity via three-dimensional porous alumina

Science.gov (United States)

Abad, Begoña; Maiz, Jon; Ruiz-Clavijo, Alejandra; Caballero-Calero, Olga; Martin-Gonzalez, Marisol

2016-01-01

Three-dimensional anodic alumina templates (3D-AAO) are an astonishing framework with open highly ordered three-dimensional skeleton structures. Since these templates are architecturally different from conventional solids or porous templates, they teem with opportunities for engineering thermal properties. By establishing the mechanisms of heat transfer in these frameworks, we aim to create materials with tailored thermal properties. The effective thermal conductivity of an empty 3D-AAO membrane was measured. As the effective medium theory was not valid to extract the skeletal thermal conductivity of 3D-AAO, a simple 3D thermal conduction model was developed, based on a mixed series and parallel thermal resistor circuit, giving a skeletal thermal conductivity value of approximately 1.25 W·m−1·K−1, which matches the value of the ordinary AAO membranes prepared from the same acid solution. The effect of different filler materials as well as the variation of the number of transversal nanochannels and the length of the 3D-AAO membrane in the effective thermal conductivity of the composite was studied. Finally, the thermal conductivity of two 3D-AAO membranes filled with cobalt and bismuth telluride was also measured, which was in good agreement with the thermal model predictions. Therefore, this work proved this structure as a powerful approach to tailor thermal properties. PMID:27934930

Arbitrary quadrature: its application in the solution of one-dimensional, planar neutron transport problems

International Nuclear Information System (INIS)

Sanchez, J.

2010-10-01

A standard numerical procedure for the solution of singular integral equations is applied to the one-dimensional transport equation for monoenergetic neutrons. As is usual in quadrature methods, the procedure yields an Eigen system whose solution provide, for the critical slab, both the eigenvalue which is proportional to the number of secondary neutrons per collision, and the density as a function of position. The results obtained with two versions of the procedure, differing only in the extent of the basic region to which they are applied, are compared with analytically derived results available for benchmarking. The procedures considered yield consistent results for the calculated neutron densities and eigenvalues. Since the one-dimensional transport kernel and its spatial moments are integrable and their integrals can be put in terms of exponential integral functions, the resulting approximations to the neutron density yield somewhat lengthy but closed, forms. These approximate expressions of the neutron density can be used to render, after they are operated on, closed-form formulas for build-up factors, extrapolation distances or angular densities or employed for other purposes that require an analytical expression of the neutron density. As an example of this latter capability, the results of the calculation of the angular density at the surface of the slab are provided. (Author)

Scattering of thermal neutron by the water molecule

International Nuclear Information System (INIS)

Rosa, L.P.

The calculation of the differenctial cross section for scattering of thermal neutrons by water, taking into account the translational, rotational and vibrational motions of the water molecule, is presented according to Nelkin' model. Some modifications are presented which have been introduced in the original method to improve the results and an application has been made to reactor physics, by calculating the thermal neutron flux in a homogenous medium containing water and absorver. Thirty thermal energy groups have been used to compute the spectra. Within the limits of error, better agreement has been obtained between theory and experiments by using a modified Nelkin kernel consisting of substituting the asymptotic formulae for the rotational and vibrational motions by more exact expressions, similar to the Buttler model for heavy water

Indirect and direct measurement of thermal neutron acceleration by inelastic scattering on the {sup 177}Lu isomer

Energy Technology Data Exchange (ETDEWEB)

Belier, G.; Roig, O.; Meot, V.; Daugas, J.M. [CEA Bruyeres-le-Chatel, Dept. de Physique Theorique et Appliquee, 91 (France); Aupiais, J.; Jutier, Ch.; Le Petit, G. [CEA Bruyeres-le-Chatel, Service de Physique Nucleaire, 91 (France). Dept. de Physique Theorique et Appliquee; Letourneau, A.; Marie, F. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee, Service de Physique Nucleaire, 91- Gif sur Yvette (France); Veyssiere, Ch. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee, Service d' Ingenierie des Systemes, 91- Gif sur Yvette (France)

2008-07-01

When neutrons interact with isomers, these isomers can de-excite and in such a reaction the outgoing neutron has an energy greater than the in-going one. This process is referred as Inelastic Neutron Acceleration or Super-elastic Scattering. Up to now this process was observed for only two nucleus, {sup 152m}Eu and {sup 180m}Hf by measuring the number of fast neutrons produced by isomeric targets irradiated with thermal neutrons. In these experiments the energies of the accelerated neutrons were not measured. This report presents an indirect measurement of inelastic neutron acceleration on {sup 177m}Lu, based on the burn-up and the radiative capture cross sections measurements. Since at thermal energies the inelastic scattering and the radiative capture are the only processes that contribute to the isomer burn-up, the inelastic cross section can be deduced from the difference between the two measured quantities. Applying this method for the {sup 177}Lu isomer with different neutron fluxes we obtained a value of (257 {+-} 50) barns (for a temperature of 323 K) and determined that there is no integral resonance for this process. In addition the radiative capture cross section on {sup 177g}Lu was measured with a much better accuracy than the accepted value. Since the acceleration cross section is quite high, a direct measurement of this process was undertaken, sending thermal neutrons and measuring the fast neutrons. The main goal now is to measure the outgoing neutron energies in order to identify the neutron transitions in the exit channel. In particular the K conservation question can be addressed by such a measurement. (author)

The OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark - Steady-state results and status

International Nuclear Information System (INIS)

Reitsma, F.; Han, J.; Ivanov, K.; Sartori, E.

2008-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated. Since only a few pebble-bed HTR experimental facilities or plant data are available the use of code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MW design and a representative set of transient cases is defined as an OECD benchmark. The scope of the benchmark is to establish a series of well-defined multi-dimensional computational benchmark problems with a common given set of cross-sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. The OECD benchmark includes steady-state and transients cases. Although the focus of the benchmark is on the modelling of the transient behaviour of the PBMR core, it was also necessary to define some steady-state cases to ensure consistency between the different approaches before results of transient cases could be compared. This paper describes the status of the benchmark project and shows the results for the three steady state exercises defined as a standalone neutronics calculation, a standalone thermal-hydraulic core calculation, and a coupled neutronics/thermal-hydraulic simulation. (authors)

Study of the Li2CO3 as thermal neutrons detector

International Nuclear Information System (INIS)

Herrera A, E.; Urena N, F.; Delfin L, A.

2003-01-01

The use every day but it frequents of the thermal neutrons in the treatment of tumours, using the neutron capture therapy technique in boron, there is generated the necessity to develop a dosimetric system that allows to evaluate in a reliable way the fluence and consequently the dose of neutrons that it is given in the tumours of the patients. One of the techniques but employees to determine the neutron fluence sub cadmic and epi cadmic in an indirect way, it is the activation of thin sheets of gold undress and covered with cadmium respectively that when being exposed to a neutron beam to the nuclear reaction 197 Au (n, γ ) 198 Au, emitting gamma radiation with an energy of 0.4118 MeV, being this, a disadvantage to be used as dosemeter. On the other hand, when exposing the lithium carbonate to a thermal neutron beam, free radicals of CO 3 that are quantified by the electron paramagnetic resonance technique are generated. This work analyzes those basic parameters that determine if those made up of Li 2 CO 3 complete with the requirements to be used as detectors and/or dosemeters of thermal neutrons. (Author)

Nuclear graphite based on coal tar pitch; behavior under neutron irradiation between 400 and 14000C

International Nuclear Information System (INIS)

Mottet, P.; Fillatre, A.; Schill, R.; Micaud, G.

1977-01-01

Two nuclear grades of coal tar pitch coke graphites have been developed and tested under neutron irradiation. The neutron irradiation induced dimensional changes between 400 and 1400 0 C, at fluences up to 1,2.10 22 n.cm -2 PHI.FG show a behavior comparable to anisotropic petroleum coke graphites. Less than 10% variation in thermal expansion, maximum decrease by a factor four in thermal conductivity, and large increase of the Young modulus have been observed

Characterization of two Japanese ancient swords through neutron imaging

International Nuclear Information System (INIS)

Grazzi, Francesco; Zoppi, Marco; Salvemini, Filomena; Kaestner, Anders; Lehmann, Eberhard; Civita, Francesco

2015-01-01

Japanese blades are culturally interesting objects both from the stylistic point of view and because of their fantastic performances. In this work, we present new results, using a non-invasive approach, concerning these peculiar artefacts. Two integer Japanese swords, pertaining to Kyoto (987-1596) and Shinto (1596-1781) periods have been analysed through neutron-imaging techniques. The experiments have been performed at the ICON beam line, operating at the spallation neutron source SINQ, Paul Scherrer Institut in Switzerland. The reconstruction of projection data into neutron tomographic slices or volumes, allowed us to identify some very peculiar characteristics, related to the forging methods and to the different thermal treatments applied to produce the cutting edge and its unique feature. (author)

Characterization of thermal neutron fields for calibration of neutron monitors in accordance with great equivalent dose environment H⁎(10)

International Nuclear Information System (INIS)

Silva, Larissa P. S. da; Silva, Felipe S.; Fonseca, Evaldo S.; Patrao, Karla C.S.; Pereira, Walsan W.

2017-01-01

The Laboratório Brasileiro de Nêutrons do Instituto de Radioproteção e Dosimetria (IRD/CNEN) has developed and built a thermal neutron flux facility to provide neutron fluence for dosimeters (Astuto, 2014). This fluency is obtained by four 16 Ci sources 241 AmBe (α, n) positioned around the channel positioned in the center of the Thermal Flow Unit (UFT). The UFT was built with blocks of paraffin with graphite addition and graphite blocks of high purity to obtain a central field with a homogeneous thermal neutron fluence for calibration purposes with the following measurements: 1.2 x 1.2 x 1.2 m 3 . The objective of this work is to characterize several points, in the thermal energy range, in terms of the equivalent ambient dose quantity H⁎(10) for calibration and irradiation of monitors neutrons

A two-step approach for the preliminary evaluation of the thermal-hydraulics and safety of the ELSY open square core design

International Nuclear Information System (INIS)

Meloni, Paride; Bandini, Giacomino; Polidori, Massimiliano; Cervone, Antonio; Manservisi, Sandro

2009-01-01

Several innovative solutions for a liquid metal fast reactor design have been investigated in the EURATOM Sixth Framework Programme and an open-assembly core design for the ELSY (European Lead-cooled System) reactor has been proposed by ENEA. The development of this new reactor, based on innovative neutronic and safety considerations, requires a new approach to the thermal-hydraulic (T/H) core design. In this paper a new two-step approach of the T/H analysis for this open-assembly core is presented and, in particular is used for the evaluation of the preliminary core design of a 1500 MW lead fast reactor with open square lattice and three fuel radial zones with different levels of enrichment. In the first step a preliminary thermal-hydraulic and safety evaluation of the core neutronic design is investigated by using a one-dimensional RELAP5 model for independent channel analysis. Then two and three-dimensional effects are taken into account by using a dedicated tool for the evaluation of assembly mixing effects. The RELAP5 model, based on pressure loss and heat transfer correlations available for heavy liquid metal flows in rod bundle, consists of completely independent assemblies and therefore it can be used for a conservative evaluation of the thermal-hydraulics of the core reactor. Due to the open-lattice configuration, the two and three-dimensional effects are important and they are taken into account by using a simplified three-dimensional numerical model of an open square lattice reactor core, developed with the purpose of analyzing the whole core behavior. The numerical simulation is performed at assembly length level taking into account the local fluctuations of turbulent viscosity and energy exchange coefficients at sub-channel level through transfer operators based on parametric coefficients. A preliminary evaluation of the mixing effects between assembly flows on the temperature field has been performed by using an average assembly turbulent viscosity

Seismically constrained two-dimentional crustal thermal structure of ...

Indian Academy of Sciences (India)

Cambay basin; P-wave velocity; heat flow; heat generation; 2-D modelling; crustal thermal structure; Mohodepth; Curie isotherm. ... This work deals with the two-dimensional thermal modelling to delineate the crustal thermal structure along a 230 km long Deep Seismic Sounding (DSS) profile in the north Cambay basin.

Geant4 Analysis of a Thermal Neutron Real-Time Imaging System

Science.gov (United States)

Datta, Arka; Hawari, Ayman I.

2017-07-01

Thermal neutron imaging is a technique for nondestructive testing providing complementary information to X-ray imaging for a wide range of applications in science and engineering. Advancement of electronic imaging systems makes it possible to obtain neutron radiographs in real time. This method requires a scintillator to convert neutrons to optical photons and a charge-coupled device (CCD) camera to detect those photons. Alongside, a well collimated beam which reduces geometrical blurriness, the use of a thin scintillator can improve the spatial resolution significantly. A representative scintillator that has been applied widely for thermal neutron imaging is 6LiF:ZnS (Ag). In this paper, a multiphysics simulation approach for designing thermal neutron imaging system is investigated. The Geant4 code is used to investigate the performance of a thermal neutron imaging system starting with a neutron source and including the production of charged particles and optical photons in the scintillator and their transport for image formation in the detector. The simulation geometry includes the neutron beam collimator and sapphire filter. The 6LiF:ZnS (Ag) scintillator is modeled along with a pixelated detector for image recording. The spatial resolution of the system was obtained as the thickness of the scintillator screen was varied between 50 and 400 μm. The results of the simulation were compared to experimental results, including measurements performed using the PULSTAR nuclear reactor imaging beam, showing good agreement. Using the established model, further examination showed that the resolution contribution of the scintillator screen is correlated with its thickness and the range of the neutron absorption reaction products (i.e., the alpha and triton particles). Consequently, thinner screens exhibit improved spatial resolution. However, this will compromise detection efficiency due to the reduced probability of neutron absorption.

Utilizing the slowing-down-time technique for benchmarking neutron thermalization in graphite

International Nuclear Information System (INIS)

Zhou, T.; Hawari, A. I.; Wehring, B. W.

2007-01-01

Graphite is the moderator/reflector in the Very High Temperature Reactor (VHTR) concept of Generation IV reactors. As a thermal reactor, the prediction of the thermal neutron spectrum in the VHTR is directly dependent on the accuracy of the thermal neutron scattering libraries of graphite. In recent years, work has been on-going to benchmark and validate neutron thermalization in 'reactor grade' graphite. Monte Carlo simulations using the MCNP5 code were used to design a pulsed neutron slowing-down-time experiment and to investigate neutron slowing down and thermalization in graphite at temperatures relevant to VHTR operation. The unique aspect of this experiment is its ability to observe the behavior of neutrons throughout an energy range extending from the source energy to energies below 0.1 eV. In its current form, the experiment is designed and implemented at the Oak Ridge Electron Linear Accelerator (ORELA). Consequently, ORELA neutron pulses are injected into a 70 cm x 70 cm x 70 cm graphite pile. A furnace system that surrounds the pile and is capable of heating the graphite to a centerline temperature of 1200 K has been designed and built. A system based on U-235 fission chambers and Li-6 scintillation detectors surrounds the pile. This system is coupled to multichannel scaling instrumentation and is designed for the detection of leakage neutrons as a function of the slowing-down-time (i.e., time after the pulse). To ensure the accuracy of the experiment, careful assessment was performed of the impact of background noise (due to room return neutrons) and pulse-to-pulse overlap on the measurement. Therefore, the entire setup is surrounded by borated polyethylene shields and the experiment is performed using a source pulse frequency of nearly 130 Hz. As the basis for the benchmark, the calculated time dependent reaction rates in the detectors (using the MCNP code and its associated ENDF-B/VI thermal neutron scattering libraries) are compared to measured

Set of thermal neutron-scattering experiments for the Weapons Neutron Research Facility

International Nuclear Information System (INIS)

Brugger, R.M.

1975-12-01

Six classes of experiments form the base of a program of thermal neutron scattering at the Weapons Neutron Research (WNR) Facility. Three classes are to determine the average microscopic positions of atoms in materials and three are to determine the microscopic vibrations of these atoms. The first three classes concern (a) powder sample neutron diffraction, (b) small angle scattering, and (c) single crystal Laue diffraction. The second three concern (d) small kappa inelastic scattering, (e) scattering surface phonon measurements, and (f) line widths. An instrument to couple with the WNR pulsed source is briefly outlined for each experiment

Design of thermal neutron beam based on an electron linear accelerator for BNCT.

Science.gov (United States)

Zolfaghari, Mona; Sedaghatizadeh, Mahmood

2016-12-01

An electron linear accelerator (Linac) can be used for boron neutron capture therapy (BNCT) by producing thermal neutron flux. In this study, we used a Varian 2300 C/D Linac and MCNPX.2.6.0 code to simulate an electron-photoneutron source for use in BNCT. In order to decelerate the produced fast neutrons from the photoneutron source, which optimize the thermal neutron flux, a beam-shaping assembly (BSA) was simulated. After simulations, a thermal neutron flux with sharp peak at the beam exit was obtained in the order of 3.09×10 8 n/cm 2 s and 6.19×10 8 n/cm 2 s for uranium and enriched uranium (10%) as electron-photoneutron sources respectively. Also, in-phantom dose analysis indicates that the simulated thermal neutron beam can be used for treatment of shallow skin melanoma in time of about 85.4 and 43.6min for uranium and enriched uranium (10%) respectively. Copyright © 2016. Published by Elsevier Ltd.

Thermal neutron equivalent doses assessment around KFUPM neutron source storage area using NTDs

Energy Technology Data Exchange (ETDEWEB)

Abu-Jarad, F.; Fazal-ur-Rehman; Al-Haddad, M.N.; Al-Jarrallah, M.I.; Nassar, R

2002-07-01

Area passive neutron dosemeters based on nuclear track detectors (NTDs) have been used for 13 days to assess accumulated low doses of thermal neutrons around neutron source storage area of the King Fahd University of Petroleum and Minerals (KFUPM). Moreover, the aim of this study is to check the effectiveness of shielding of the storage area. NTDs were mounted with the boron converter on their surface as one compressed unit. The converter is a lithium tetraborate (Li{sub 2}B{sub 4}O{sub 7}) layer for thermal neutron detection via {sup 10}B(N,{alpha}){sup 7}Li and {sup 6}Li(n,{alpha}){sup 3}H nuclear reactions. The area passive dosemeters were installed on 26 different locations around the source storage area and adjacent rooms. The calibration factor for NTD-based area passive neutron dosemeters was found to be 8.3 alpha tracks.cm{sup -2}.{mu}Sv{sup -1} using active snoopy neutron dosemeters in the KFUPM neutron irradiation facility. The results show the variation of accumulated dose with locations around the storage area. The range of dose rates varied from as low as 40 nSv.h{sup -1} up to 11 {mu}Sv.h{sup -1}. The study indicates that the area passive neutron dosemeter was able to detect accumulated doses as low as 40 nSv.h{sup -1}, which could not be detected with the available active neutron dosemeters. The results of the study also indicate that an additional shielding is required to bring the dose rates down to background level. The present investigation suggests extending this study to find the contribution of doses from fast neutrons around the neutron source storage area using NTDs through proton recoil. The significance of this passive technique is that it is highly sensitive and does not require any electronics or power supplies, as is the case in active systems. (author)

Thermalization of a two-dimensional photonic gas in a `white wall' photon box

Science.gov (United States)

Klaers, Jan; Vewinger, Frank; Weitz, Martin

2010-07-01

Bose-Einstein condensation, the macroscopic accumulation of bosonic particles in the energetic ground state below a critical temperature, has been demonstrated in several physical systems. The perhaps best known example of a bosonic gas, blackbody radiation, however exhibits no Bose-Einstein condensation at low temperatures. Instead of collectively occupying the lowest energy mode, the photons disappear in the cavity walls when the temperature is lowered-corresponding to a vanishing chemical potential. Here we report on evidence for a thermalized two-dimensional photon gas with a freely adjustable chemical potential. Our experiment is based on a dye-filled optical microresonator, acting as a `white wall' box for photons. Thermalization is achieved in a photon-number-conserving way by photon scattering off the dye molecules, and the cavity mirrors provide both an effective photon mass and a confining potential-key prerequisites for the Bose-Einstein condensation of photons. As a striking example of the unusual system properties, we demonstrate a yet unobserved light concentration effect into the centre of the confining potential, an effect with prospects for increasing the efficiency of diffuse solar light collection.

TRIDENT: a two-dimensional, multigroup, triangular mesh discrete ordinates, explicit neutron transport code

International Nuclear Information System (INIS)

Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.

1977-03-01

TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided

Long-term thermal two- and three-dimensional analysis of roller compacted concrete dams supported by monitoring verification

Energy Technology Data Exchange (ETDEWEB)

Kuzmanovic, V.; Savic, L. [Belgrade Univ. (Serbia). Faculty of Civil Engineering; Stefanakos, J. [National Technical Univ. of Athens (Greece). Dept. of Water Resources and Environmental Engineering

2010-04-15

This study investigated the long-term thermal-field evolution of roller compacted concrete (RCC) dams. Thermal computational analyses of the dams are needed as a result of the layer-based construction technologies used to build the dams. Two-dimensional (2-D) and 3-D unsteady phased models of the RCC dams were used to determine the time evolution of thermal field in a dam based on the Platanovryssi dam in Greece. The finite element method (FEM) was used to account for the dam geometry, different types of concrete used; actual initial and boundary conditions; the thermal and mechanical properties of the dam as a function of aging and temperature; and the RCC construction technology. The influence of all the parameters on the thermal behaviour of the RCC gravity dam was analyzed. Results of the study showed that the 2-D model accurately described the RCC dam thermal field. The thermal behaviour of the dam was influenced primarily by the thermal properties of the mixture and the boundary conditions. Variations of layer thickness did not significantly influence the temperature field. 18 refs., 3 tabs., 10 figs.

Facility at CIRUS reactor for thermal neutron induced prompt γ-ray spectroscopic studies

International Nuclear Information System (INIS)

Biswas, D.C.; Danu, L.S.; Mukhopadhyay, S.; Kinage, L.A.; Prashanth, P.N.; Goswami, A.; Sahu, A.K.; Shaikh, A.M.; Chatterjee, A.; Choudhury, R.K.; Kailas, S.

2013-01-01

A facility for prompt γ-ray spectroscopic studies using thermal neutrons from a radial beam line of Canada India Research Utility Services (CIRUS) reactor, Bhabha Atomic Research Centre (BARC), has been developed. To carry out on-line spectroscopy experiments, two clover germanium detectors were used for the measurement of prompt γ rays. For the first time, the prompt γ–γ coincidence technique has been used to study the thermal neutron induced fission fragment spectroscopy (FFS) in 235 U(n th , f). Using this facility, experiments have also been carried out for on-line γ-ray spectroscopic studies in 113 Cd(n th , γ) reaction

Absence of storage effects on radiation damage after thermal neutron irradiation of dry rice seeds

Energy Technology Data Exchange (ETDEWEB)

Kowyama, Y. [Mie Univ., Tsu (Japan); Saito, M.; Kawase, T.

1987-09-15

Storage effects on dry rice seeds equilibrated to 6.8% moisture content were examined after irradiation with X-rays of 5, 10, 20 and 40 kR and with thermal neutrons of 2.1, 4.2, 6.3 and 8.4×10{sup 13}N{sub th}/cm{sup 2}. Reduction in root growth was estimated from dose response curves after storage periods of 1 hr to 21 days. The longer the storage period, the greater enhancement of radiation damages in X-irradiated seeds. There were two components in the storage effect, i. e., a rapid increase of radiosensitivity within the first 24 hr and a slow increase up to 21 days. An almost complete absence of a storage effect was observed after thermal neutron exposure, in spite of considerably high radioactivities of the induced nuclides, {sup 56}Mn, {sup 42}K and {sup 24}Na, which were detected from gamma-ray spectrometry of the irradiated seeds. The present results suggest that the contributions of gamma-rays from the activated nuclides and of inherent contaminating gamma-rays are little or negligible against the neutron-induced damage, and that the main radiobiological effects of thermal neutrons are ascribed to in situ radiations, i, e., heavy particles resulting from neutron-capture reaction of atom. A mechanism underlying the absence of storage effect after thermal neutron irradiation was briefly discussed on the basis of radical formation and decay. (author)

Investigation of Response of Several Neutron Surveymeters by a DT Neutron Generator

International Nuclear Information System (INIS)

Kim, Sang In; Jang, In Su; Kim, Jang Lyul; Lee, Jung IL; Kim, Bong Hwan

2012-01-01

Several neutron measuring devices were tested under the neutron fields characterized with two distinct kinds of thermal and fast neutron spectrum. These neutron fields were constructed by the mixing of both thermal neutron fields and fast neutron fields. The thermal neutron field was constructed using by a graphite pile with eight AmBe neutron sources. The fast neutron field of 14 MeV was made by a DT neutron generator. In order to change the fraction of fast neutron fluence rate in each neutron fields, a neutron generator was placed in the thermal neutron field at 50 cm and 150 cm from the reference position. The polyethylene neutron collimator was used to make moderated 14 MeV neutron field. These neutron spectra were measured by using a Bonner sphere system with an LiI scintillator, and dosimetric quantities delivered to neutron surveymeters were determined from these measurement results.

Bibliography for thermal neutron scattering

International Nuclear Information System (INIS)

Sakamoto, Masanobu; Chihara, Junzo; Gotoh, Yorio; Kadotani, Hiroyuki; Sekiya, Tamotsu.

1979-09-01

Bibliographic references are given for measurements, calculations, reviews and basic studies of thermal neutron scattering and dynamical properties of condensed matter. This is the sixth edition covering 3,326 articles collected up to 1978. The edition being the final issue of the present bibliography series, a forthcoming edition will be published in a new form of bibliography. (author)

Experimental characterization of HOTNES: A new thermal neutron facility with large homogeneity area

Energy Technology Data Exchange (ETDEWEB)

Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Sperduti, A. [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); ENEA C.R. Frascati, via E. Fermi n. 45, 00044 Frascati, Roma (Italy); Pietropaolo, A.; Pillon, M. [ENEA C.R. Frascati, via E. Fermi n. 45, 00044 Frascati, Roma (Italy); Pola, A. [Politecnico di Milano, Dipartimento di Energia, via La Masa 34, 20156 Milano (Italy); INFN–Milano, Via Celoria 16, 20133 Milano (Italy); Gómez-Ros, J.M. [INFN–LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

2017-01-21

A new thermal neutron irradiation facility, called HOTNES (HOmogeneous Thermal NEutron Source), was established in the framework of a collaboration between INFN-LNF and ENEA-Frascati. HOTNES is a polyethylene assembly, with about 70 cmx70 cm square section and 100 cm height, including a large, cylindrical cavity with diameter 30 cm and height 70 cm. The facility is supplied by a {sup 241}Am-B source located at the bottom of this cavity. The facility was designed in such a way that the iso-thermal-fluence surfaces, characterizing the irradiation volume, coincide with planes parallel to the cavity bottom. The thermal fluence rate across a given isofluence plane is as uniform as 1% on a disk with 30 cm diameter. Thermal fluence rate values from about 700 cm{sup −2} s{sup −1} to 1000 cm{sup −2} s{sup −1} can be achieved. The facility design, previously optimized by Monte Carlo simulation, was experimentally verified. The following techniques were used: gold activation foils to assess the thermal fluence rate, semiconductor-based active detector for mapping the irradiation volume, and Bonner Sphere Spectrometer to determine the complete neutron spectrum. HOTNES is expected to be attractive for the scientific community involved in neutron metrology, neutron dosimetry and neutron detector testing.

THERMAL: A routine designed to calculate neutron thermal scattering

International Nuclear Information System (INIS)

Cullen, D.E.

1995-01-01

THERMAL is designed to calculate neutron thermal scattering that is isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the center of mass system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy

Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

Science.gov (United States)

Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

2018-01-01

The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

Histological and Physiological Alterations Induced by Thermal Neutron Fluxes in Male Swiss Albino Mice

International Nuclear Information System (INIS)

Alzergy, A.A.; Emara, N.M.; Abd El-Latif, A.A.; El-Saady, S.M.M.; Emara, N.M.; Abd El-Latif, A.A.

2010-01-01

This work was performed to investigate the biological effects of different thermal neutron fluxes (0.27x10 8 , 0.52X10 8 , 1.089X10 8 , 2.16X10 8 and 4.32X10 8 ) on liver and kidney of male mice using neutron irradiation cell with Ra-Be(α,n) 3 mCi neutron source Leybold (55930). Exposed to various fluxes of thermal neutron induced a dramatic alterations in hepatic and renal functions as indicated by biochemical estimation of several parameters (bilirubin, SGT, and alkaline phosphate .Urea , total protein, and albumin) and confirmed by histological examinations Thermal neutron exposure induces marked increase in the serum activities of total bilirubin, alanine amino transaminase (ALT or GPT), and alkaline phosphate, whereas, urea, total protein and albumin showed marked decline as compared to control group. The physiological changes induced in thermal neutron fluxes dependent manner. Histopathological results revealed mild to severe type of necrosis, and degenerative changes in liver and kidney of male mice exposed to thermal neutron fluxes. Also it was found that the histopathological alterations induced in thermal neutron fluxes dependent manner. It was found that exposed to thermal neutron fluxes irradiation plays prominent role in the development of the physiological alterations in male Swiss albino mice. The Former up normalities as a result of the sequence events followed interaction of radiation with the former biological mater (liver and kidney) of male Swiss albino mice, which are, physical, physicochemical, chemical, and biological stages.

Manufacturing of thermal neutron sensor using pMOS

International Nuclear Information System (INIS)

Lee, Nam Ho; Kim, Seung Ho

2005-05-01

A pMOSFET sensor having a Gadolinium converter has been invented successfully as a slow neutron sensor that is sensitive to neutron energy down to 0.025 eV. The Gd layer converts low energy neutrons to ionizing radiation of which the amount is proportional to neutron dose. Ionising radiation from neutron reactions changes the charge state of the gate oxide of the pMOSFET. The Gd-pMOSFETs were tested at a neutron beam port of HANARO research reactor and a 60 CO irradiation facility to investigate slow neutron response and gamma response, respectively. The voltage change was proportional to the accumulated slow neutron dose. The results from Gd coupled MOSFET neutron dosemeters shows an excellent sensitivity (15 - 16mV/cGy) and linearity to thermal neutrons with negligible background contamination. The results demonstrate the outstanding performance of the Gd coupled MOSFET neutron dosemeters clearly. The Gd-pMOSFET can also be used in a mixed radiation field by subtracting the voltage change of a pMOSFET without Gd from that of the Gd-pMOSFET

The OECD/NEA/NSC PBMR 400 MW coupled neutronics thermal hydraulics transient benchmark: transient results - 290

International Nuclear Information System (INIS)

Strydom, G.; Reitsma, F.; Ngeleka, P.T.; Ivanov, K.N.

2010-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of well defined multi-dimensional computational benchmark problems with a common given set of cross sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants' results are compared using a statistical method and possible areas of future code improvement are identified. (authors)

The band gap variation of a two dimensional binary locally resonant structure in thermal environment

Directory of Open Access Journals (Sweden)

Zhen Li

2017-01-01

Full Text Available In this study, the numerical investigation of thermal effect on band gap dynamical characteristic for a two-dimensional binary structure composed of aluminum plate periodically filled with nitrile rubber cylinder is presented. Initially, the band gap of the binary structure variation trend with increasing temperature is studied by taking the softening effect of thermal stress into account. A breakthrough is made which found the band gap being narrower and shifting to lower frequency in thermal environment. The complete band gap which in higher frequency is more sensitive to temperature that it disappears with temperature increasing. Then some new transformed models are created by changing the height of nitrile rubber cylinder from 1mm to 7mm. Simulations show that transformed model can produce a wider band gap (either flexure or complete band gap. A proper forbidden gap of elastic wave can be utilized in thermal environment although both flexure and complete band gaps become narrower with temperature. Besides that, there is a zero-frequency flat band appearing in the first flexure band, and it becomes broader with temperature increasing. The band gap width decreases trend in thermal environment, as well as the wider band gap induced by the transformed model with higher nitrile rubber cylinder is useful for the design and application of phononic crystal structures in thermal environment.

Coupled thermal-hydraulic and neutronic simulations of Phenix control rod withdrawal tests with SIMMER-IV

International Nuclear Information System (INIS)

Kriventsev, Vladimir; Gabrielli, Fabrizio; Rineiski, Andrei

2014-01-01

The “end-of-life” tests performed in the Phenix reactor before its final shutdown in 2009, in particular the Control Rod (CR) withdrawal experiments provide an excellent opportunity for the validation and verification of the reactor physics computer codes and modeling approaches. SIMMER-IV, a modern three-dimensional reactor safety code, has been recently employed at Karlsruhe Institute of Technology (KIT) for simulating Phenix experiments in the framework of a benchmark exercise organized under the IAEA project. In this paper, we report and discuss main results obtained with SIMMER-IV at KIT. Particular attention is devoted to the coupling features of thermal-hydraulics and neutronics and their mutual influences. The reactor reactivity, power and neutron flux distributions calculated with SIMMER-IV are in good agreement both with experimental results and with calculations with advanced neutronics codes, such as ERANOS, while the CR reactivity worth is overestimated due to neglecting heterogeneity effects. Because of its multi-physics capabilities SIMMER also calculates the temperature distributions which are in a good agreement with the experimental test results. In this work we describe the improvements in SIMMER neutronics model by employing a correction that is based on the results of cell calculations performed with ERANOS. The study confirms that the 3D SIMMER-IV code can accurately predict major fast reactor neutronics and thermal hydraulic parameters, provided that a special treatment is employed for CR modeling. The results of calculations are analyzed in frames of SIMMER-IV validation and verification assessment. (author)

ESR-dosimetry in thermal and epithermal neutron fields for application in boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Schmitz, Tobias

2016-01-22

Dosimetry is essential for every form of radiotherapy. In Boron Neutron Capture Therapy (BNCT) mixed neutron and gamma fields have to be considered. Dose is deposited in different neutron interactions with elements in the penetrated tissue and by gamma particles, which are always part of a neutron field. The therapeutic dose in BNCT is deposited by densely ionising particles, originating from the fragmentation of the isotope boron-10 after capture of a thermal neutron. Despite being investigated for decades, dosimetry in neutron beams or fields for BNCT remains complex, due to the variety in type and energy of the secondary particles. Today usually ionisation chambers combined with metal foils are used. The applied techniques require extensive effort and are time consuming, while the resulting uncertainties remain high. Consequently, the investigation of more effective techniques or alternative dosimeters is an important field of research. In this work the possibilities of ESR-dosimeters in those fields have been investigated. Certain materials, such as alanine, generate stable radicals upon irradiation. Using Electron Spin Resonance (ESR) spectrometry the amount of radicals, which is proportional to absorbed dose, can be quantified. Different ESR detector materials have been irradiated in the thermal neutron field of the research reactor TRIGA research reactor in Mainz, Germany, with five setups, generating different secondary particle spectra. Further irradiations have been conducted in two epithermal neutron beams. The detector response, however, strongly depends on the dose depositing particle type and energy. It is hence necessary to accompany measurements by computational modelling and simulation. In this work the Monte Carlo code FLUKA was used to calculate absorbed doses and dose components. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using amorphous track models. For the simulation, detailed models of

Thermal field theory in a layer: Applications of thermal field theory methods to the propagation of photons in a two-dimensional electron sheet

International Nuclear Information System (INIS)

Nieves, Jose F.

2010-01-01

We apply the thermal field theory methods to study the propagation of photons in a plasma layer, that is a plasma in which the electrons are confined to a two-dimensional plane sheet. We calculate the photon self-energy and determine the appropriate expression for the photon propagator in such a medium, from which the properties of the propagating modes are obtained. The formulas for the photon dispersion relations and polarization vectors are derived explicitly in some detail for some simple cases of the thermal distributions of the charged particle gas, and appropriate formulas that are applicable in more general situations are also given.

Neutron detector development at Brookhaven

International Nuclear Information System (INIS)

Yu, B.; Harder, J.A.; Mead, J.A.; Radeka, V.; Schaknowski, N.A.; Smith, G.C.

2003-01-01

Two-dimensional thermal neutron detectors have been the subject of research and development at Brookhaven for over 20 years. Based primarily on multi-wire chambers filled with a gas mixture containing 3 He, these detectors have been used in wide-ranging studies of molecular biology and material science samples. At each phase of development, experimenters have sought improvements in key parameters such as position resolution, counting rate, efficiency, solid-angle coverage and stability. A suite of detectors has been developed with sensitive areas ranging from 5x5 to 50x50 cm 2 . These devices incorporate low-noise-position readout and the best position resolution for thermal neutron gas detectors. Recent developments include a 1.5 mx20 cm detector containing multiple segments with continuously sensitive readout, and detectors with unity gain for ultra-high rate capability and long-term stability

Transmission of germanium poly- and monocrystals for thermal neutrons at different temperatures

International Nuclear Information System (INIS)

Adib, M.; Abdel-Kawy, A.; Eid, Y.; Maayouf, R.M.; Abbas, Y.; Habib, N.; Kilany, M.; Ashry, A.

1987-01-01

Neutron cross-sections of germanium poly- and monocrystals were measured with two time-of-flight and two double-axis crystal spectrometers. The results were analyzed using the single-level Breit-Wigner formula. The coherent scattering amplitude was determined from the Bragg reflections observed in the cross-section of a polycrystal and the analysis of the neutron diffraction pattern. The incoherent and the thermal diffuse scattering cross-section were estimated from the analysis of the total cross-section data obtained for a monocrystal at different temperatures in the energy range 2 meV to 1 eV. (orig./HP) [de

Transmission of germanium poly- and monocrystals for thermal neutrons at different temperatures

Energy Technology Data Exchange (ETDEWEB)

Adib, M.; Abdel-Kawy, A.; Eid, Y.; Maayouf, R.M.; Abbas, Y.; Habib, N.; Kilany, M.; Ashry, A.

Neutron cross-sections of germanium poly- and monocrystals were measured with two time-of-flight and two double-axis crystal spectrometers. The results were analyzed using the single-level Breit-Wigner formula. The coherent scattering amplitude was determined from the Bragg reflections observed in the cross-section of a polycrystal and the analysis of the neutron diffraction pattern. The incoherent and the thermal diffuse scattering cross-section were estimated from the analysis of the total cross-section data obtained for a monocrystal at different temperatures in the energy range 2 meV to 1 eV.

Low dimensional equivalence of core neutronics model and its application to transient analysis

International Nuclear Information System (INIS)

Song Hongbing; Zhao Fuyu

2015-01-01

Three-dimensional coupled neutronics thermal-hydraulics reactor analysis is time consuming and occupies huge memory. A one-dimensional model is preferable than the three one in nuclear system analysis, control system design and load following. In this paper, a corewide three dimensional to one dimensional equivalent method has been developed. On the basis of this method 1D axial few groups constants were obtained. The equivalent cross sections were calculated by general spatial homogenization while the transverse buckling was computed through an equivalence based on the 3D flux conservation. Three steady test cases were performed on one dimensional finite difference code ODTAC and the results were compared with TRIVAC-5. The comparison shows that the one dimensional axial power distribution computed by ODTAC correlates well with the three dimensional results calculated by TRIVAC-5. In this study, DRAGON-4 was used to generate the few-group constants of fuel assemblies and the reflector few-group parameters were calculated by WIMS-D4. These collapsed few-group constants were tabulated in a database sorted in ascending order of fuel temperature, coolant temperature and concentration of boric acid. Trilinear interpolation was adopted in cross sections feedback during the transient analysis. In this paper, G1 rod drop accident (RDA) and G1 rod ejection accident (REA) were performed on ODTAC and the computation results were consistent of the physical rules. (author)

Thermal neutron flux distribution in the ET R R-1 reactor core as experimentally measured and theoretically calculated by the code triton

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

Thermal neutron flux distributions that were measured earlier at the ET-R R-1 reactor are compared with those calculated by the three dimensional diffusion code Triton. This comparison was made for the horizontal and vertical flux distributions. The horizontal thermal flux distributions considered in this comparison were along the core diagonals at two planes of different heights from core bottom, where one at a level passing through the control rod at core center and the other at a level below this control rod. In the meantime all the control rods were taken into consideration. The effect of the existence of a water cavity inside the core as well as the influence of the control rods on the thermal flux are illustrated in this work. The vertical thermal flux distributions considered in the comparison were at two positions in core namely; one along the core height the horizontal reactor power distribution along the core height and the horizontal reactor power distribution along the core diagonal as calculated by the code Triton are also given this work. 8 figs., 1 tab.

Development of the variety for resistance against bacterial leaf-blight in rice with thermal neutrons

International Nuclear Information System (INIS)

Nakai, Hirokazu

1990-01-01

In search for the development of genes for resistance against bacterial leaf-blight in rice, thermal neutrons generated from the Research Reactor at the Kyoto University have been applied to the breeding. In this paper, the developmental outcome is described, and a potential application of thermal neutrons for breeding the variety of resistance against bacterial leaf-blight in rice is reviewed. When thermal neutrons were delivered to the rice, the ratio of absorbed doses by B-10, which is contained in a small quantity in the plant, was found to be larger than expected. This implies characteristic effects of thermal neutrons on the plant. When boric acid was incorporated into the plant before irradiation, the effect of thermal neutrons per irradiation time was considered to become great. The frequency of mutations for resistance was significantly higher by thermal neutron, as compared with that induced by other mutagens, such as gamma radiation, ethylene-imine, ethyl-methane-sulfonate, and nitroso-methyl-urea. Genetic analysis of mutants for resistance revealed recessive genes and polygenes. Finally, the application of thermal neutrons and other radiations would contribute greatly to a resolution of serious pollution problems in global food and environment. (N.K.)

Neutron irradiation effect of thermally-sensitized stainless steels

Energy Technology Data Exchange (ETDEWEB)

Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

1998-03-01

Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

Using MCNP-4C code for design of the thermal neutron beam for neutron radiography at the MNSR

International Nuclear Information System (INIS)

Shaaban, I.

2009-11-01

Studies were carried out for determination of the parameters of a thermal neutron beam at the MNSR reactor (MNSR-30 kW) for neutron radiography in the vertical beam port by using the MCNP-4C (Monte Carlo Neutron - Photon transport). Thermal, epithermal and fast neutron energy ranges were selected as 10 keV respectively. To produce a good neutron beam in terms of intensity and quality, several materials Lead (Pb), Bismuth (Bi), Borated polyethelyene and Alumina Oxide (Al 2 O 3 ) were used as neutron and photon filters. Based on the current design, the L/D of the facility ranges between 125, 110 and 90. The thermal neutron flux at the beam exit is 1.436x10 5 n/cm2 .s ,1.843x10 5 n/cm2 .s and 2.845x10 5 n/cm2 .s respectively, middots with a Cd-ratio of ∼ 2.829, 2.766, 3.191 for the L/D = 125, 110, 90 respectively. The estimated values for gamma doses are 6.705x10 -2 Rem/h and 1.275x10 -1 Rem/h and 2.678x10 -1 Rem/ h with bismuth. The divergent angle of the collimator is 1.348 degree - 2.021 degree. Such neutron beams, if built into the Syrian MNSR reactor, could support the application of NRG in Syria. (author)

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

Science.gov (United States)

Bowman, Charles D.

1992-01-01

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

Science.gov (United States)

Bowman, C.D.

1992-11-03

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

Optimization of thermal neutron shield concrete mixture using artificial neural network

Energy Technology Data Exchange (ETDEWEB)

Yadollahi, A. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box: 1983963113, Tehran (Iran, Islamic Republic of); Nazemi, E., E-mail: nazemi.ehsan@yahoo.com [Young Researchers and Elite Club, Kermanshah Branch, Islamic Azad University, Kermanshah (Iran, Islamic Republic of); Zolfaghari, A. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box: 1983963113, Tehran (Iran, Islamic Republic of); Ajorloo, A.M. [Water and Environmental Engineering Department, Shahid Beheshti University, P.O. Box: 167651719, Tehran (Iran, Islamic Republic of)

2016-08-15

Highlights: • Colemanite was used in fabricating of thermal neutron shield concrete. • The Taguchi method was implemented to obtain the data set required for training the ANN. • Trained ANN predicted quality characteristics of thermal neutron shield. - Abstract: Colemanite is the most convenient boron mineral which has been widely used in construction of radiation shielding concrete in order to improve the capture of thermal neutrons. But utilization of Colemanite in radiation shielding concrete has a deleterious effect on both physical and mechanical properties. In the present work, Taguchi method and artificial neural network (ANN) were employed to find an optimal mixture of Colemanite based concrete in order to improve the boron content of concrete and increase thermal neutron absorption without violating the standards for physical and mechanical properties. Using Taguchi method for experimental design, 27 concrete samples with different mixtures were fabricated and tested. Water/cement ratio, cement quantity, volume fraction of Colemanite aggregate and silica fume quantity were selected as control factors, and compressive strength, ultrasonic pulse velocity and thermal neutron transmission ratio were considered as the quality responses. Obtained data from 27 experiments were used to train 3 ANNs. Four control factors were utilized as the inputs of 3 ANNs and 3 quality responses were used as the outputs, separately (each ANN for one quality response). After training the ANNs, 1024 different mixtures with different quality responses were predicted. At the final, optimum mixture was obtained among the predicted different mixtures. Results demonstrated that the optimal mixture of thermal neutron shielding concrete has a water–cement ratio of 0.38, cement content of 400 kg/m{sup 3}, a volume fraction Colemanite aggregate of 50% and silica fume–cement ratio of 0.15.

Optimization of thermal neutron shield concrete mixture using artificial neural network

International Nuclear Information System (INIS)

Yadollahi, A.; Nazemi, E.; Zolfaghari, A.; Ajorloo, A.M.

2016-01-01

Highlights: • Colemanite was used in fabricating of thermal neutron shield concrete. • The Taguchi method was implemented to obtain the data set required for training the ANN. • Trained ANN predicted quality characteristics of thermal neutron shield. - Abstract: Colemanite is the most convenient boron mineral which has been widely used in construction of radiation shielding concrete in order to improve the capture of thermal neutrons. But utilization of Colemanite in radiation shielding concrete has a deleterious effect on both physical and mechanical properties. In the present work, Taguchi method and artificial neural network (ANN) were employed to find an optimal mixture of Colemanite based concrete in order to improve the boron content of concrete and increase thermal neutron absorption without violating the standards for physical and mechanical properties. Using Taguchi method for experimental design, 27 concrete samples with different mixtures were fabricated and tested. Water/cement ratio, cement quantity, volume fraction of Colemanite aggregate and silica fume quantity were selected as control factors, and compressive strength, ultrasonic pulse velocity and thermal neutron transmission ratio were considered as the quality responses. Obtained data from 27 experiments were used to train 3 ANNs. Four control factors were utilized as the inputs of 3 ANNs and 3 quality responses were used as the outputs, separately (each ANN for one quality response). After training the ANNs, 1024 different mixtures with different quality responses were predicted. At the final, optimum mixture was obtained among the predicted different mixtures. Results demonstrated that the optimal mixture of thermal neutron shielding concrete has a water–cement ratio of 0.38, cement content of 400 kg/m 3 , a volume fraction Colemanite aggregate of 50% and silica fume–cement ratio of 0.15.

Two-dimensional cross-section sensitivity and uncertainty analysis of the LBM experience at LOTUS

International Nuclear Information System (INIS)

Davidson, J.W.; Dudziak, D.J.; Pelloni, S.; Stepanek, J.