Radioisotope production at the IFMIF-DONES facility

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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Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

Nuclear Technology and Radiation Protection ◽

10.2298/ntrp0702003a ◽

2007 ◽

Vol 22 (2) ◽

pp. 3-9

Author(s):

Ivana Avramovic ◽

Milan Pesic

Keyword(s):

Neutron Flux ◽

Flux Density ◽

Neutron Source ◽

Deuteron Beam ◽

Target Volume ◽

Neutron Generation ◽

Research Facility ◽

Critical Research ◽

Neutron Flux Density ◽

Computer Codes

The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF) being developed over the last couple of years at the Vinca Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step) are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step). The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield) for the selected sub-critical core cells for both beams have also been presented in this paper.

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Statistical Analysis of Tritium Breeding Ratio Deviations in the DEMO Due to Nuclear Data Uncertainties

Applied Sciences ◽

10.3390/app11115234 ◽

2021 ◽

Vol 11 (11) ◽

pp. 5234

Author(s):

Jin Hun Park ◽

Pavel Pereslavtsev ◽

Alexandre Konobeev ◽

Christian Wegmann

Keyword(s):

Monte Carlo ◽

Fusion Reactor ◽

Nuclear Data ◽

Data File ◽

Nuclear Model ◽

Model Parameters ◽

Tritium Breeding ◽

Tritium Breeding Ratio ◽

Nuclear Data Library ◽

Breeding Ratio

For the stable and self-sufficient functioning of the DEMO fusion reactor, one of the most important parameters that must be demonstrated is the Tritium Breeding Ratio (TBR). The reliable assessment of the TBR with safety margins is a matter of fusion reactor viability. The uncertainty of the TBR in the neutronic simulations includes many different aspects such as the uncertainty due to the simplification of the geometry models used, the uncertainty of the reactor layout and the uncertainty introduced due to neutronic calculations. The last one can be reduced by applying high fidelity Monte Carlo simulations for TBR estimations. Nevertheless, these calculations have inherent statistical errors controlled by the number of neutron histories, straightforward for a quantity such as that of TBR underlying errors due to nuclear data uncertainties. In fact, every evaluated nuclear data file involved in the MCNP calculations can be replaced with the set of the random data files representing the particular deviation of the nuclear model parameters, each of them being correct and valid for applications. To account for the uncertainty of the nuclear model parameters introduced in the evaluated data file, a total Monte Carlo (TMC) method can be used to analyze the uncertainty of TBR owing to the nuclear data used for calculations. To this end, two 3D fully heterogeneous geometry models of the helium cooled pebble bed (HCPB) and water cooled lithium lead (WCLL) European DEMOs were utilized for the calculations of the TBR. The TMC calculations were performed, making use of the TENDL-2017 nuclear data library random files with high enough statistics providing a well-resolved Gaussian distribution of the TBR value. The assessment was done for the estimation of the TBR uncertainty due to the nuclear data for entire material compositions and for separate materials: structural, breeder and neutron multipliers. The overall TBR uncertainty for the nuclear data was estimated to be 3~4% for the HCPB and WCLL DEMOs, respectively.

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Summary of the International Workshop on Nuclear Data for Fusion Reactor Technology, San Diego, California, May 3–6, 1993

Fusion Technology ◽

10.13182/fst94-a30257 ◽

1994 ◽

Vol 25 (4) ◽

pp. 502-502

Author(s):

Frederick M. Mann

Keyword(s):

San Diego ◽

Fusion Reactor ◽

International Workshop ◽

Nuclear Data ◽

Reactor Technology

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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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Design of a Target and Moderator at the Los Alamos Spallation Radiation Effects Facility (LASREF) as a Neutron Source for Fusion Reactor Materials Development

Reactor Dosimetry ◽

10.1520/stp15174s ◽

2009 ◽

pp. 773-773-9

Author(s):

PD Ferguson ◽

GE Mueller ◽

WF Sommer ◽

EH Farnum

Keyword(s):

Neutron Source ◽

Radiation Effects ◽

Fusion Reactor ◽

Los Alamos ◽

Materials Development ◽

Reactor Materials

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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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