1. Neutron capture cross section measurements of 238U, 241Am and 243Am at n_TOF
Proposal to the ISOLDE and Neutron Time-of-Flight Committee
Spokespersons: Daniel Cano Ott, Frank Gunsing
Technical coordinator: Vasilis Vlachoudis

2. High level Nuclear waste
Man-made nucleosynthesis: the 235U nuclear fuel cycle
High level
Nuclear waste
b--decay
(n,g)
a emission

3. Computational design tools
Nuclear data for the transmutation of nuclear waste
Transmutation of the Minor Actinides by (n,f) and (n,g) in new nuclear systems, thus reducing:
radiotoxicity inventory 1/100
cooling time 1/1000
Computational design tools
Need of accurate& reliable NUCLEAR DATA

4. Isotopes which have become relevant due to their increased presence in the fuels for the new generation of reactors. This applies in particular to the systems aiming at the Pu and Minor Actinides (MA) recycling: fast reactors and ADS, but also Gen III(+) reactors with a high Pu load.
The proposal of fast systems, which has enhanced the relevance of the region from 1eV to several MeV for all isotopes and, very especially, for high mass actinides.
New fuel cycles with multi-recycling of actinides that could drive to new levels of accumulation of minor isotopes, thus generating additional risks and/or costs at several stages of the fuel cycle. One of the most important aspects is the propagation of the uncertainties as the number of irradiation cycles increases.
New requirements on the level of precision. In the nuclear industry as everywhere else, the computer simulations should help minimizing (but not replacing) the need for small scale and costly experimental demonstrators. This new role of the simulation tools will require a higher level of precision that can only be achieved if the precision of initial basic nuclear data is previously improved.
Better assessment of the uncertainties on the data and on the derived magnitudes. This requires a more realistic evaluation of the cross section uncertainties and their correlations between different energies and different channels. Only in this way the estimations from the simulations will profit from the enhanced basic data precision and from the constraints set by the integral experiments.

5. 238U as the future nuclear fuel
The main energy producing reaction in commercial nuclear reactors is the fission of 235U in a thermal spectrum.
The fresh nuclear fuels contain 2%-4% of 235U, being the rest 238U, which is about 100 times more abundant in nature. There are studies which conclude that there exist 235U reserves (at a reasonable price) for 50 – 500 years, depending on the growth and future use of nuclear energy.
The long term sustainability of nuclear fission electricity production will rely on the exploitation of the breeding reactions (in fast reactors):
238U → 239U → 239Np → 239Pu
232Th → 233Th → 233Pa → 233U

6. Recent sensitivity analysis using state of art covariance data evaluations (BOLNA).
NEA/WPEC subgroup 26
M. Salvatores and R. Jacqmin (Eds),
NEA/WPEC-26.
ISBN
The uncertainties in the neutron capture cross section of 238U need to be lowered to 2-3%, depending on the fast reactor system.

7. The number of measurements on the 238U(n,γ) is much larger, over 25 datasets in subregions of the resolved resonance region, and a few less in the unresolved energy region are available in the EXFOR database. Inconsistencies are still present for the capture cross section up to 25 keV.
The goal of proposing a new measurement is the reduction of the actual uncertainty in the cross section to a value below 2% in the range of a few eV up to hundred keV. This will result into a new standard.
Such a challenge is only achievable through an ongoing European effort, which consists in a series of measurements combining different facilities, experimental techniques and analysis methodologies:
Measurements at n_TOF and IRMM Geel/GELINA.
Total Absorption Spectrometry and low neutron sensitivity C6D6 detectors.
Neutron capture and transmission measurements.
Use of (at least) two different analysis codes: SAMMY and REFIT (and eventually CONRAD).

8. Fast Spectrum Transmutation Scheme
The role of 241Am and 243Am in the transmutation scenarios
Fast Spectrum Transmutation Scheme

9. Relevant isotopes in the multi-recycling scenario
Report of the Numerical results from the Evaluation of the nuclear data sensitivities, Priority list and table of required accuracies for nuclear data. E. Gonzalez-Romero (Ed), NUDATRA Deliverable D5.11 from IP-Eurotrans
T= Transmutation efficiency
DH= Decay Heat load
N = Neutron emission
R = Radiotoxicity

10. The 241Am and 243Am data
The goal of this proposal is to improve subtantially the nuclear data accuracy on the 241Am and 243Am neutron capture cross sections, as demanded by the design transmutation + multi-recycling strategies.

11. Neutron cross sections
Reaction products
n (atoms/cm2)
I (neutrons/cm2/s)
Detecting capture reactions means to detect the subsequent EM cascade. One has to do that with a counting efficiency which depends on a simple way on the EM de-excitation pattern:
-Total Energy Detectors like the C6D6, whose efficiency is NEARLY proportional to the energy of the EM cascade.
-A Total Absorption Calorimeter, whose efficiency DOES NOT depend on the EM de-excitation pattern.
It is not straightforward to fin Plutonium samples!!
GUIÑO ALOS TEORICOS !!!

12. The TAC detection system for (n,g) reactions
neutron
Detect capture reactions using radioactive/LOW MASS samples.

13. High counting rate in resonances
The n_TOF Total Absorption Calorimeter
The n_TOF Collaboration built the Total Absorption Calorimeter based on the FZK’s:
BaF2 crystals (fast time, reasonable energy resolution, low neutron sensitivity),
95% of 4p and 15 cm crystal thickness (high total absorption efficiency),
40 modules (segmentation),
with important upgrades required at n_TOF (long flight path  resonances):
2 TB/day
High counting rate in resonances
H(n,n)
H(n,g)
10B(n,a)
Ba(n,g)
(n,n) g
neutron
NEUTRON SENSIT. APPEARS IN RESONANCES -> neutron shielding by moderation of neutrons and captre in H and B-10
DAQ: una imagen vale mas que mil palabras. CONTAR BIEN -> DIGITAL. PROBLEMA: 2TB/DAY -> NO PROBLEM AT CERN
Hay que decir que en FZK son 5-10 cm, los neutrones llegan mas tarde que lso gammas. En nuestro caso los neutrons coniciden ene l tiempo con los gammas. CONLUSION: mejorar blindaje .-
Long fligfht paths -> resonances -> neutrons cattering background and hgh counting rates (PIONERING digital DAQ)
Esquemita de neutron absorbido.
En una fuentetan intensa, tenos que ser muy precisos a la horade contar. COMO? Guardando toda la info cada 2 ns. .
En FZK funciono perfectamente, pero es un challenge acoplarlo a TOF con spallation.

14. The n_TOF Total Absorption Calorimeter
A new shielding + sample container (ISO 2919) design will lead to a 10 times lower neutron background in the calorimeter.

15. The C6D6 (n,γ) setup
Home made carbon fibre C6D6 detectors with a neutron sensitivity of 10-4.

16. The 232Th (n,γ) measurement at n_TOF
F.Gunsing, Nuc. Data. Conf.-2007
G. Aerts et al. (n_TOF Collaboration), Phys. Rev. C 73, (2006)
Resolved Resonance Region
Unresolved Resonance Region
Mass (232Th) = g
Diameter = 1.5 cm
Purity = 99.5%
High peak n flux intensity reduce the radioactive background
+ high resolution
-> larger RRR +
Small resonances

17. High accuracy measurements wit the TAC: 237Np
n_TOF capture +
GELINA transmission =
One of the best measur. made in Europe.
C. Guerrero et al. (n_TOF Collaboration), Proc. Int. Conf. Nuc. Data for Sci. and Tech. 2007, Nice.

19. Appendices

20. Partial Uncertainties
237Np capture yield: normalization and uncertainties
Two possibilities for normalization of the capture yield:
Normalization from the Saturated Resonance Method applied to 197Au.
The uncertainty associated to a possible misalignment can be calculated from the information neutron beam profile.
2. Normalization to reliable transmission data, preserving the total cross section.
The data of Gressier et al. at IRMM/GELINA are reliable and cover the complete energy range of this work.
Partial Uncertainties
Total
uncertainty
Norm. e
Alignment
Mass
F(En)
197Au at 4.9 eV 3%
2.5% 1% 2%
<5.4%
237Np(n,tot) -
<3.6%
COMENTAR ALGO DE NORNLIZAR A TRANSMISION: supone más trabahjo pero es mas parecido al o que hacen los evaluadores: RELIABILITY
The definitive results correspond to the combined analysis of capture and transmission.

21. Validation with simple γ-ray cascades: calibration sources
The detection efficiency of the TAC
Validation with simple γ-ray cascades: calibration sources

22. Photo-absorption (GDR)
Generation of realistic capture cascades requires the knowledge of the complete nuclear level scheme and transition probabilities.
Capture state
Statistical region:
Described by the nuclear level density and photon strength functions.
Transition probability given by:
TXL(Ei,Ef) = fXL(Eg) (Ef,I,)XL={ E1,M1, E2}
E*~Sn+En
Eg (MeV)
Photo-absorption (GDR)
TAC (n,g) cascades
The Photon Strength Functions are not very well known, but there exists experimental information.
b-decay
Known level scheme (ENSDF):
Energy, spin and parity, transition probability, electron conversion coefficients, ...
The experimental at high excitation energy is scarce or NON EXISTENT.
TAC USEFULL FOR NUCLEAR STRUCTURE (details in the manuscript)
Ground state

23. The reliability of the detection efficiency calculated by means of MC simulations relies on the reproducibility of all the available experimental results:
Reproducing all distribution (Esum,mcr) is a PROOF of the correctness of the MC method
Reproducing the experimental data for all combinations of Esum and mcr is a proof of the correctness of the MC simulation