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NAÏADE 1 Concrete Benchmark (60cm)

 1. Name of Experiment: 
    Fontenay aux Roses 60cm Fission Concrete Benchmarks in NAÏADE 1 followed by pure
    thermal neutron benchmarks in light water and in concrete.

 2. Purpose and Phenomena Tested: 
    The main purpose is the determination of the fission neutron transport in concrete for
    penetration up to 100 cm for fast neutron measurements and up to about 120 cm for thermal
    neutrons measurements. Considering the discrepancy between thermal neutron measurements
    and calculations, we propose benchmarks in water and in concrete using a pure thermal neutron

 3. Description of the Source and Experimental Configuration: 
    The NAÏADE 1 facility, located in Fontenay aux Roses, was composed of a thermal or fission
    neutron source reaching a large experimental area (27 cubic meters) in which various
    shielding mock-ups were placed in order to validate nuclear constants and computer
    programs. The thermal neutron source is the direct collimated beam coming from the core of
    the ZOE heavy water reactor surrounded by a graphite reactor. This beam is collimated with
    two sizes (60 cm and 40 cm) and can be used as a neutron source directly. For obtaining
    a fission source a uranium plate was interposed between the collimator plate and the
    experimental area. The plate is 1 square meter and the thickness of the fissile part is
    2 cm. It consists of 9 square tiles, 0.333 m along the side, made of natural uranium,
    clad in 1mm of aluminum. Behind the fission plate, which generates fission neutrons, there
    was the experimental area in which an aluminum tank was placed. A boral screen separated
    the fission plate from the experimental area to avoid thermal neutron backscattering.
    In the concrete experiments, the concrete block thickness was 200 cm, and its section was
    200 cm×200 cm enclosed in a steel frame 1cm thick and surrounded by a thick concrete shield
    to avoid neutron bypassing. The chemical compositions of two concrete samples, S2 and S5
    are given.
    The absolute fission neutron source distribution is determined by Monte Carlo
    calculation (TRIPOLI 4) using the thermal neutron flux measurements (Mn55) for the source
    term. This determination takes into account all fissions (U235 and U238) provided by the
    neutron diffusions in the converter itself, in structures and in the mock-up (a sub-critical
    system with a source of thermal neutrons). In this experiment, the diaphragm diameter was
    60 cm.
    Despite all precautions taken into account in the fission benchmark interpretation using
    the TRIPOLI 4 program and ENDF/B/VI R4, a discrepancy is found between the measurement and
    calculation results in the thermal energy range for all penetration distances. The present
    benchmark also contains measurements and interpretations for 2 experiments using the pure
    thermal neutron source: 1] experimental area filled by light water (to improve the pure
    thermal neutron beam intensity and its weak intermediate neutron background noise) and 2]
    with the concrete block (to improve the total macroscopic absorption cross-section of
    the concrete)
    In all parts of these benchmarks, the background noise is calculated and compared with
    available measurements.

 4. Measurement System and Uncertainties: 
    The used detectors were:
    	1. Phosphorus dosimeters (P31(n,p) reaction)
    	2. Rhodium dosimeters (Rh103(n,n') reaction)
    	3. Silicon diodes (Wigner effect)
    	4. Gold deposit Au197 covered with cadmium ((n,?) reaction)
    	5. Indium 115 covered with cadmium ((n,?) reaction)
    	6. Mn55 foils covered with cadmium ((n,?) reaction)
    All dosimeters were calibrated in well-known fluxes depending on their characteristics:
    Maxwellian thermal flux at 27°C in a reference block, fission spectrum with correction
    taking into account the diffusion effects altering slightly the pure fission spectrum,
    constant flux per unit of lethargy.
    All experimental results were given with a ZOE reactor power equal to 100 kW. The power
    stability and reproducibility were checked using a boron fluoride ?-compensated
    ionization chamber and Mn55 dosimeters placed on a rod located before the fission plate.
    The precision of the fission plate power varies from 1% when the measurements are made
    close together in time (a few days) to 5% when the measurements are separated by several
    months. During the dosimeter irradiation, the observed stability is dP/P = 0.5%.
    The dosimeter position uncertainty estimated by the experimental team itself is +/- 0.1 cm.

 5. Description of Results and Analysis: 
    All results are expressed in conventional fluxes. (Equivalent fission flux, flux per unit
    of lethargy, equivalent thermal flux at 2200 m/s). The corresponding mean cross-section or
    integral of resonance are given. It is possible to express all calculated results in
    reaction rates without ambiguity. The as-measured experimental results are given on the
    converter axis for several distances. A calibrated dosimeter reassessment resulting from
    the nuclear data improvements was made recently (2003-2007) and published.
    The interpretation using the French CEA Monte-Carlo code TRIPOLI 4 was made on these
    concrete benchmarks. A background noise evaluation was also made using at the same time
    TRIPOLI 4 calculations and fast and epithermal neutron flux measurements without a converter
    plate. The corresponding input and output data sets are included (see following table
    in section 10).

 6. Special Features: 

 7. Author/Organizer 
    Experiment and Analysis:
	M. Lott, P. Pepin, L. Bourdet, G. Cabaret, J. Capsie, M. Dubor, M. Hot, C. Goulet
	CEA (French Atomic Energy Commission), DPA/DEP/SEPP
	92260 Fontenay-aux-Roses, France

    Compilation of data for Sinbad and experiment interpretation using TRIPOLI 4:
	J.C. Nimal
	CEA Centre de Saclay DEN/DM2S/SERMA/
	91191 Gif-sur-Yvette Cedex, France

    Reviewer of compilated data:
	I. Kodeli
	OECD/NEA, 12 bd des Iles, 92130 Issy-les-Moulineaux, France

 8. Availability: 

 9. References:
    [1] Lott M., Pepin P., Bourdet L., Cabaret G., Capsie J., Dubor M., Hot M., Goulet C.
    (décembre 1970), Étude expérimentale de l’atténuation des neutrons dans différents
    matériaux de protection à l’aide du dispositif NAÏADE 1 du réacteur ZOE; Note CEA 1386.
    [2] Nimal J.C. (December 2011), New interpretation of the NAÏADE 1 Experiments
    – Part 1 The Iron and Graphite Experiments, NEA/NSC/DOC(2005)15/REV1.
	Nimal J.C. (décembre 2011), Nouvelles interprétations des expériences NAÏADE 1
    – Partie 1 Expérience sur le fer et le graphite, NEA/NSC/DOC(2005)15/REV1.
    [3] Dulieu P. (janvier 1965), Utilisation pratique du détecteur de dommages au silicium,
    Note CEA 514.
	De Cosnac B., Dulieu P., Le Ralle J.C., Manent G. (4-8 mars 1968), Mesure des flux
    de neutrons rapides au moyen de diodes en silicium sensibles aux dommages, « Colloque
    d’Electronique Nucléaire et Radioprotection, Tome 1 », Toulouse, France.
    [4] Nimal J.C. (December 2011), New interpretation of the NAÏADE 1 Experiments
    – Part 2 Fission Neutron Propagation in Light Water, source diameter 60 cm,
	Nimal J.C. (décembre 2011), Nouvelles interprétations des expériences NAÏADE 1
    – Partie 2 Propagation des neutrons de fission dans l’eau légère, diamètre de
    la source 60 cm, NEA/NSC/DOC(2006)24/REV1.
    [5] Both J. P., Lee Y.K., Mazzolo A., Petit O., Peneliau Y., Roesslinger B.,
    Soldevila M. (September 2003), TRIPOLI 4 – A Three Dimensional Polykinetic Particle
    Transport Monte-Carlo Code, SNA’2003, Paris.
    	Both J.P., Mazzolo A., Peneliau Y., Petit O., Roesslinger B. (2003),
    Notice d’utilisation du code TRIPOLI-4.3 : code de transport de particules par
    la méthode de Monte-Carlo, Rapport CEA-R-6043.
    	Both, J.P., Mazzolo A., Peneliau Y., Petit O., Roesslinger B. (2003),
    User manual for version 4.3 of the TRIPOLI 4 Monte-Carlo Method Particle
    Transport Computer Code, Rapport CEA-R-6044.
    [6] JEF report 14. Table of Simple Integral Neutron Cross Section Data from JEF-2.2,
    ENDF/B-VI, JENDL-3.2, BROND-2 and CENDL-2; OECD/NEA, Paris.
    [7] Case K. M., de Hoffmann F., Plazczek G. (June 1953), Introduction to the theory
    of neutron diffusion, Volume I; Los Alamos Scientific Laboratory, Los Alamos, New Mexico.
    [8] Nimal J.C., Réévaluation des réponses des diodes au silicium et du détecteur
    Au197(n,gamma). Application aux benchmarks relatifs à l’atténuation des neutrons
    de fission dans le fer, le graphite et l’eau légère. - Partie I, OCDE/AEN, Paris.
    	Nimal J.C., Reassessment of the silicon diode and Au197(n,gamma) dosimeters.
    Application to the fission neutron attenuation in iron, graphite and light water
    - Part I, OECD/NEA, Paris.
    [9] Radiation Shielding and Dosimetry Experiments Database (SINBAD);
    OECD/NEA, Paris. //www.oecd-nea.org/science/shielding/sinbad/sinbadis.htm.

 10. Data and Format: 

     The following three items are available in the Data set SINBAD (NEA-1517/087):
	- Introduction, summary and NAÏADE 1 facility description
	(from introduction up to PART A),
	- Expression of the experimental results, experiment reassessment (PART B),
	- Converter plate power determination (PART C and PART D).

     Experiments and interpretations on concrete mock-up with a fission source and
     on light water and concrete with a pure thermal source (PART H).

     Four TRIPOLI 4 input files (.data) and four TRIPOLI 4 output files (.data.res),
     referring to the direct signals (file name beginning by " naiade ", see table below).

     Four TRIPOLI 4 input files (.data) and four TRIPOLI 4 output files (.data.res),
     referring to the background noise (file name beginning by " bdf ", see table below).
     Filename     	 	   Mat     	Source 
     ------------------- 	   ----- 	------------- 
     naiadebetons2e_2007.data	   S2		fission
     naiadebetons2e_2007.data.res  S2		fission
     naiadebetons5e_2007.data	   S5		fission
     naiadebetons5e_2007.data.res  S5		fission
     naiadebetonths2_2007.data	   S2	 	thermal
     naiadebetonths2_2007.data.res S2		thermal
     naiadeeauth_2007.data	   H2O		thermal
     naiadeeauth_2007.data.res	   H2O		thermal
     bdfbetons2e_2007.data	   S2		noise
     bdfbetons2e_2007.data.res	   S2		noise
     bdfbetons5e_2007.data	   S5		noise
     bdfbetons5e_2007.data.res	   S5		noise
     bdfbetonths2e_2007.data	   S2		noise
     bdfbetonths2e_2007.data.res   S2		noise
     bdfnaiadeeauth_2007.data	   H2O		noise
     bdfnaiadeeauth_2007.data.res  H2O		noise
     naiade-c.htm		Abstract
     Part H - EN.pdf      	Description of Experiment
     Part H - FR.pdf      	Description of Experiment (in French)

     Note : In the column "file name", the digits 2 or 5 refer to the concrete
     analysed compositions S2 or S5 ; the column entitled "source" refers to the source type
     defined in the input data : "fission" source with uranium plate, pure "thermal" neutron
     source with neither uranium plate nor boral screen, "noise" for the background noise
     with uranium plate (for fission source case) or without this last one
     for thermal neutron source case).

     * Reference number 07_007 attributed twice by mistake

SINBAD Benchmark Generation Date: 1/2012
SINBAD Benchmark Last Update: 1/2012