Angular Distribution of Neutrons Around Thick Beryllium Target of Accelerator-Based 9Be(d, n) Neutron Source

Abstract An experimental work and simulations were carried out to determine the angular distributions of neutrons and yields of the 9Be(d, n) reaction over all angular range (360 deg) on a thick beryllium target as an accelerator-based neutron source at incident-deuteron energy 13.6 MeV. The neutron activation method was used in the experimental part using aluminum and iron foils as detectors to calculate the neutron flux. The Monte Carlo neutral-particles code (MCNP5) was used to demonstrate and simulate the neutron distribution, also to understand and compare with the experimental results. The neutron energy spectrum was computed using the projection angular-momentum coupled evaporation code PACE4 (LISE++) and the spectrum was adopted in MCNP5 code. Two experimental ways were used, one with beryllium target and another one without the beryllium target, to evaluate the neutron flux emitted only by the beryllium target. Typical computational results were presented and are compared with the previous experimental data to evaluate the computing model as well as the characteristics of emitted neutrons produced by the 9Be(d, n) reaction with a thick Be-target. Moreover, the results can be used to optimize the shielding and collimating system for neutron therapy.

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Double Layer Collimator for BNCT Neutron Source Based on 30 MeV Cyclotron

Indonesian Journal of Physics and Nuclear Applications ◽

10.24246/ijpna.v2i3.124-127 ◽

2017 ◽

Vol 2 (3) ◽

pp. 124

Author(s):

Bilalodin Bilalodin ◽

Kusminarto Kusminarto ◽

Arief Hermanto ◽

Yohannes Sardjono ◽

Sunardi Sunardi

Keyword(s):

Neutron Flux ◽

Fast Neutron ◽

Neutron Source ◽

Double Layer ◽

Thermal Neutron Flux ◽

Epithermal Neutron ◽

Epithermal Neutron Flux ◽

Beryllium Target ◽

Neutron Filter ◽

Neutron Current

A research of design of double layer collimator using 9Be(p,n) neutron source has been conducted. The research objective is to design a double layer collimator to obtain neutron sources that are compliant with the IAEA standards. The approach to the design of double layer collimator used the MCNPX code. From the research, it was found that the optimum dimensions of a beryllium target are 0.01 mm in length and 9.5 cm in radius. Collimator consists of a D2O and Al moderator, Pb and Ni as a reflector, and Cd and Fe as a thermal and fast neutron filter. The gamma filter used Bi and Pb. The quality neutron beams emitted from the double layer collimator is specified by five parameters: epithermal neutron flux 1 ×109 n/cm2s; fast neutron dose per epithermal neutron flux 5 ×1013 Gy cm2s; gamma dose per epithermal neutron flux 1×1013 Gy cm2s; ratio of the thermal neutron flux of epithermal neutron flux 0; and the ratio of epithermal neutron current to total epithermal neutron 0.54.

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Damage Calculation for First Wall Submitted to High Neutron Flux in a Tokamak

MRS Proceedings ◽

10.1557/opl.2015.124 ◽

2015 ◽

Vol 1769 ◽

Author(s):

C.E. Velasquez ◽

M. A. F. Veloso ◽

A. L. Costa ◽

C. Pereira

Keyword(s):

Neutron Flux ◽

Energy Deposition ◽

Reaction Rates ◽

Main Reaction ◽

First Wall ◽

Tokamak Reactor ◽

Mcnp5 Code ◽

Displacement Per Atom ◽

High Neutron ◽

High Neutron Flux

ABSTRACTThe displacement per atom (dpa) has been a specific issue to evaluate the damage in the first wall of the Tokamak. Two different first wall materials were evaluated. In this study, MCNP5 code was used to obtain the neutron flux, the energy deposition and the main reaction rates, on the inboard and outboard first wall. The damage calculations were performed by the SPECTER code using the neutronic parameters obtained by MCNP5. The Tokamak reactor modeled has similar dimensions to the ITER. Tungsten and beryllium alloys were simulated on the outboard first wall. The results indicate which material has a higher resistance to be damage and dpa values for the analyzed material.

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Optimization of the testing volumes with respect to neutron flux levels in the two-target high flux D-Li neutron source for the international fusion materials irradiation facility

10.1109/fusion.1989.102309 ◽

2003 ◽

Author(s):

W.P. Kelleher ◽

G.L. Versamis

Keyword(s):

Neutron Flux ◽

Neutron Source ◽

High Flux ◽

Fusion Materials

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A setup with a mechanical chopper for monochromating neutrons and shortening a neutron flux from a pulsed neutron source

Instruments and Experimental Techniques ◽

10.1134/s0020441206050022 ◽

2006 ◽

Vol 49 (5) ◽

pp. 617-623

Author(s):

Zh. V. Mezentseva ◽

A. P. Sirotin ◽

Yu. V. Grigor’ev ◽

H. Faikov-Stanczyk

Keyword(s):

Neutron Flux ◽

Neutron Source ◽

Pulsed Neutron ◽

Pulsed Neutron Source ◽

Mechanical Chopper

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Neutron Distribution in Multilayered Media for a Point Fast Neutron Source

Soviet Advances in Nuclear Geophysics ◽

10.1007/978-1-4899-4947-9_10 ◽

1965 ◽

pp. 124-128

Author(s):

O. A. Barsukov ◽

V. S. Avzyanov

Keyword(s):

Fast Neutron ◽

Neutron Source ◽

Neutron Distribution ◽

Multilayered Media

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Interferences in instrumental neutron activation analysis by threshold reactions and uranium fission for miniature neutron source reactor

Radiochimica Acta ◽

10.1524/ract.2013.2064 ◽

2013 ◽

Vol 101 (9) ◽

pp. 601-606

Author(s):

M. Wasim

Keyword(s):

Neutron Activation Analysis ◽

Instrumental Neutron Activation Analysis ◽

Activation Analysis ◽

Neutron Activation ◽

Neutron Flux ◽

Neutron Source ◽

Fission Products ◽

Correction Factors ◽

Uranium Fission ◽

Instrumental Neutron Activation

Summary Miniature neutron source reactors (MNSR) are known for their stable neutron flux characteristics and are mostly employed for neutron activation analysis (NAA). Interfering reactions are sometimes observed in instrumental neutron activation analysis (INAA). Failure to correct for these interferences produces significant systematic positive errors. This paper provides correction factors for the interferences caused by the threshold reactions and fission products of 235U. These factors were calculated by using the experimentally determined thermal, epithermal and fast neutron flux and epithermal neutron flux shape factor and the nuclear data from the literature using the Høgdahl convention. Correction factors were calculated for (n, p) and (n, α) reactions for the most commonly observed radionuclides in INAA. Similarly, correction factors for uranium fission were calculated for 9 elements (Ce, Ba, La, Mo, Nd, Pd, Ru, Sm and Zr). The correction factors were validated by analyzing different materials. A comparison of uranium fission factors with those published in the literature showed a good agreement except for 97Zr, 99Mo and 131Ba which is due to difference in the flux characteristics. In general, these factors can be used with confidence.

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Radioisotope production at the IFMIF-DONES facility

EPJ Web of Conferences ◽

10.1051/epjconf/202023923001 ◽

2020 ◽

Vol 239 ◽

pp. 23001

Author(s):

Javier Praena ◽

Francisco Garcia-Infantes ◽

Rafael Rivera ◽

Laura Fernandez-Maza ◽

Fernando Arias de Saavedra ◽

...

Keyword(s):

Neutron Flux ◽

Neutron Source ◽

Fusion Reactor ◽

Deuteron Beam ◽

Nuclear Data ◽

Important Application ◽

Medium Size ◽

Radioisotope Production ◽

Fusion Technology ◽

Research Infrastructures

The International Fusion Materials Irradiation Facility - Demo Oriented NEutron Source (IFMIF-DONES) is a single-sited novel Research Infrastructure for testing, validation and qualification of the materials to be used in a fusion reactor. Recently, IFMIF-DONES has been declared of interest by ESFRI (European Strategy Forum on Research Infrastructures) and its European host city would be Granada (Spain). In spite the first and most important application of IFMIF-DONES related to fusion technology, the unprecedented neutron flux available could be exploited without modifying the routine operation of IFMIF-DONES. Thus, it is already planned an experimental hall for a complementary program with neutrons. Also, a complementary program on the use of the deuteron beam could help IFMIF-DONES to be more sustainable. In the present work, we study radioisotope production with deuterons of 177Lu. The results show the viability of IFMIF-DONES for such production in terms of the needs of a territory of small-medium size. Also the study suggests that new nuclear data at higher deuteron energies are mandatory for an accurate study in this field.

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