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M. Bielewicz, S. Kilim, E. Strugalska-Gola, M. Szuta , A. Wojciechowski ,

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M. Bielewicz, S. Kilim, E. Strugalska-Gola, M. Szuta , A. Wojciechowski ,

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  1. The Tenth session of the AER Working Group F-”Spent Fuel Transmutation” and INPRO IAEA Phase 2 Collaborative Project „Meeting Energy Needs in a period of Raw Materials Insufficiency during the 21’st Century” 25-28 March 2008 , at the Konferencni centrum AV CR – zamek Liblice 61 277 32 Bysice, Czech Republic.The neutron converter at a horizontal channel of research reactor MARIA for transmutation of fission products and incineration of minor actinides. M. Bielewicz, S. Kilim, E. Strugalska-Gola, M. Szuta, A. Wojciechowski, Institute of Atomic Energy, Otwock-Swierk 05-400, Poland G. De Cargouet Nuclear Physics Institute AS CR PRI, Rez near Prague, Czech Republic

  2. The neutron converter at a horizontal channel of research reactor MARIA for transmutation of fission products and incineration of minor actinides.. • Outline • Introduction • Specification of the Stand (neutron converter and neutron island)at the Horizontal Channel of the MARIA Research Reactor for the Studies of Transmutation of Minor Actinides and Fission Products. • Calculations for the converter and neutron island using probabilistic Monte Carlo methodology. • - Neutron energy range of interest for the transmutation of long living fission products and minor actinides • - Geometry simulation of the converter and the neutron island. • - Calculations of neutron flux distribution in the neutron island experimental channels for the neutron energy range from 1 eV to 10 keV. • - Calculations of neutron flux distribution in the converter for several characteristic points.

  3. Introduction • A stand at the horizontal channel of the MARIA research reactor for the studies of fission products transmutation and minor actinides incineration is a continuation of the scientific program „Energy plus Transmutation“carried by the Laboratory of High Energy, JINR, Dubna, Russia , which we are involved in since several years. • Because of a high cost, the Nuclotron accelerator is started up two times per year for two or three weeks. • An idea of the stand at the horizontal channel of the research reactor “MARIA” arose in order to intensify investigation of the transmutation processes. • We assume that the investigation of the transmutation processes we will continue in cooperation with the international group of scientists within the project “Energy plus Transmutation”.

  4. Specification of the Stand at the Horizontal Channel of the MARIA Research Reactor for the Studies of Transmutation of Minor Actinides and Fission Products. • Description of the stand – geometry. • The simplified geometry for verifying calculations of the concept (idea) of facility for studying experimentally the transmutation of fission products and minor actinides is presented . • The general view of the first configuration (set-up) – both the converter and the neutron island (lead or graphite block enabling to obtain neutron flux for resonance transmutation) is shown in Fig. beside.

  5. Specification of the Stand at the Horizontal Channel of the MARIA Research Reactor for the Studies of Transmutation of Minor Actinides and Fission Products – cont. • Description of material data. • The converter will be realised by using fuel rods of the EK-10 type fuel (see beside Fig.) placed so that the rods will be directly seen by the thermal neutron flux of the horizontal channel. • The fuel element EK-10 type is characterised: • - material – dispersion of UO2 and Mg, • -enrichment - 10% • -cladding – aluminium • -active length – 495 mm. • -Diameter of fuel sample – 7 mm • The natural uranium rods of 30 cm length, 2.72 cm diameter and 2.8735 kg weight are hermetically sealed in an aluminium cladding.

  6. Specification of the Stand at the Horizontal Channel of the MARIA Research Reactor for the Studies of Transmutation of Minor Actinides and Fission Products – cont. • The external neutron source • In general, the output thermal neutron flux at horizontal channels is equal to 3 – 5 x 109 cm-2s-1. • Thermal neutrons are the dominating component in the neutron spectrum. • Contribution of the epithermal and fast neutrons were equal to 9.7 % and 4.2 % of the total flux density respectively

  7. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology.Neutron energy range of interest for the transmutation of long living fission products and minor actinides • The potential radio-toxicity of fission products can be neglected after 250 years. • However the potential radio-toxicity of actinides remain yet high after million years. • This is clearly demonstrated on the Fig. beside.

  8. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Neutron energy range of interest for the transmutation of long living fission products and minor actinides • The following fission products Tc-99 and I-129 constitute 95 % of total activity of the long lived fission products. • salts of these two elements are soluble in water and could contaminate the biosphere • The long leaving iodine 129I after neutron capturing becomes an iodine 130I which decays to the stable xenon 130Xe according to the reaction: • 129I ( T1/2 = 1.57107 y) + n 130I (T1/2= 12.4 h) 130Xe (stable)

  9. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Neutron energy range of interest for the transmutation of long living fission products and minor actinides • In fact, the neutron absorption cross-sections in function of energy for the fission products 99Tc and 129I have resonance complex structure of cross sections. • The resonance region lies in the neutron energy range from 1 eV to 10 keV. This range is very important for the effective transmutation. • This is clearly seen on the figs.of 99Tc(upper) and 129I(lower) presented beside.

  10. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Neutron energy range of interest for the transmutation of long living fission products and minor actinides • The problem is that the actinide waste, which consists of neptunium and higher atomic number elements except plutonium, is thought not to transmute well in a thermal flux of the typical commercial power reactor (1014 n/cm2 s.) • However, for high enough thermal neutron flux (higher than 1014 n/cm2 s) the effective neptunium cross section for fission increases in function of thermal neutron flux. • Fig. of neptunium fission cross section beside shows it.

  11. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Neutron energy range of interest for the transmutation of long living fission products and minor actinides • Since we are not able to reach such an intense thermal neutron flux in such a simple facility as the considered above converter and in view of the inefficiency of actinides incineration for a low intensity thermal neutron flux, we focus our attention on fast flux zone of the facility, where the fission probability is more favourable. • This is clearly seen on the figs.of 237Np(upper) and 241Am(lower) fission neutron crosssection presented beside.

  12. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Neutron energy range of interest for the transmutation of long living fission products and minor actinides • Also fig. of 243Am fission neutron cross section versus energy presented beside supports it. • Finally, applying the simple facility at the reactor MARIA, the neutron range of interest for the transmutation of LLFP is the range from 1 eV to 10 keV and for incineration of MA is the range of fast neutrons. • For research of LLFP transmutation the neutron island may be used and of MA incineration the converter experimental channels can be used.

  13. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Geometry simulation of the converter and the neutron island. • Basing on the specification of the Stand at MARIA Research Reactor presented above in the work, a geometry approximation of the facility was performed using the methodology of MCNP4 version. • Defining each material with its density and composition we obtained the converter simulation presented in Fig. beside.

  14. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Geometry simulation of the converter and the neutron island - cont. • In Fig. beside (upper) is presented the general view of the stand – converter placed on the lead island (version 1) or on the graphite island (version 2). • In Fig. beside (lower) is presented the general view of the stand – converter with shifted central fuel rod placed on the lead island (version 3) or on the graphite island (version 4)

  15. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Calculations of neutron flux distribution in the neutron island experimental channels for the neutron energy range from 1 eV to 10 keV. • Calculations for version 1 – converter placed on the lead island. • The effective multiplication factor keff is estimated to be equal to 0.20196 with a standard deviation of 0.00131. An average number of neutrons per source particle is equal to 2.3489E+00. • Axial distribution of resonance neutron flux (1eV-10kV) (upper fig.) and of total neutron flux (lower fig.).

  16. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Calculations of neutron flux distribution in the neutron island experimental channels for the neutron energy range from 1 eV to 10 keV – cont. • Calculations for version 2 – converter placed on the graphite island. • The effective multiplication factor keff is estimated to be equal to 0.20566 with a standard deviation of 0.00040. An average number of neutrons per source particle is equal to 2.3538E+00. • Axial distribution of resonance neutron flux (1eV-10kV) (upper fig.) and of total neutron flux (lower fig.).

  17. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Calculations of neutron flux distribution in the neutron island experimental channels for the neutron energy range from 1 eV to 10 keV – cont. • Calculations for version 3 – converter with shifted central fuel rod placed on the lead island • The average number of neutron per source particle is equal to 2.9204E+00. • Axial distribution of resonance neutron flux (1eV-10kV) (upper fig.) and of total neutron flux (lower fig.).

  18. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Calculations of neutron flux distribution in the neutron island experimental channels for the neutron energy range from 1 eV to 10 keV – cont. • Calculations for version 4 – converter with shifted central fuel rod placed on the graphite island. • The average number of neutron per source particle is equal to 2.9322E+00. • Axial distribution of resonance neutron flux (1eV-10kV) (upper fig.) and of total neutron flux (lower fig.).

  19. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Calculations of neutron flux distribution in the neutron island experimental channels for the neutron energy range from 1 eV to 10 keV – cont. • Comparison of neutron flux distribution calculations in the experimental channel at the position (0,-15) for the resonance neutron energy range from 1 eV to 10 keV of the four versions is presented. • Number of simulation 106.

  20. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Calculations of neutron flux distribution in the converter for several characteristic points • For the version 1. • The chosen points of the neutron converter are presented in the Fig. beside.

  21. Calculations for the converter and neutron island using probabilistic Monte Carlo methodology - cont.Calculations of neutron flux distribution in the converter for several characteristic points – cont. • Calculation results. • For the version 1. • Neutron distribution in function of neutron energy for the chosen points of the neutron converterin the Fig, beside. • For the set-up it was tracked about 7.2 millions of particles.

  22. Thank you for the attention.

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