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NEA-1553 SINBAD FUSION--.

SINBAD FUSION, Neutronics Benchmark Experiments

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1. NAME:  SINBAD FUSION, Neutronics Benchmark Experiments.
NEA-1553/26
SINBAD-ILL-FE. University of Illinois Iron Sphere Benchmark (1975).

NEA-1553/27
SINBAD-SB5-FUS. ORNL 14-MeV Neutron Stainless-Steel/Borated Polyethylene Slab Experiment (1979).

NEA-1553/40
SINBAD-MEPhI. MEPhI empty slits streaming experiment (1995-1998).

NEA-1553/41
SINBAD-KANT. KANT Spherical Shell Transmission Experiment on Beryllium ('KANT' stands for 'Karlsruhe Neutron Transmission Experiment').

NEA-1553/43
SINBAD-LI-BLANKET. Juelich Neutronics Lithium Metal Blanket Experiment (1976 - 1984).

NEA-1553/45
SINBAD-FNS-BENCHM. Collection of experimental data for fusion neutronics benchmark (1983-1991).

NEA-1553/46
SINBAD-TUD-FE. TUD Iron Slab Experiment.

NEA-1553/47
SINBAD-TUD-FNG-W. TUD/FNG measurement of neutron and photon flux spectra in the FNG tungsten assembly (2002).

NEA-1553/48
SINBAD-OKTAVIAN/NI. Osaka Nickel Sphere Benchmark Experiment (OKTAVIAN) (1983).

NEA-1553/50
SINBAD-OKTAVIAN/FE. Osaka Iron Sphere Benchmark Experiment (OKTAVIAN) (1983).

NEA-1553/52
SINBAD-OKTAVIAN/SI. Leakage Neutron and Gamma Spectra from 40 and 60 cm diameter Silicon Sphere Pile With 14 MeV Neutrons (March 1987).

NEA-1553/53
SINBAD-OKTAVIAN-W. Leakage Neutron and Gamma Spectra from 40 cm diameter Tungsten Sphere  Pile With 14 MeV Neutrons (July 1987).

NEA-1553/54
SINBAD-FNG-BLKT. FNG Neutronics Bulk SS Shield Experiment (1995).

NEA-1553/55
SINBAD-FNG-DOSE-RATE. FNG-ITER Dose Rate Experiment (2000-2001).

NEA-1553/56
SINBAD-FNG-SIC. FNG Benchmark Experiment on Silicon Carbide (SiC) (2001).

NEA-1553/57
SINBAD-FNG-SS. SS Bulk Shield Benchmark Experiment at FNG/ENEA, FENDL Benchmark for IAEA/NDS, 1989.

NEA-1553/58
SINBAD-FNG-STREAMING. FNG-ITER NEUTRON STREAMING EXPERIMENT (1997-1998).

NEA-1553/59
SINBAD-FNG-W. FNG Benchmark Experiment on Tungsten (2001).

NEA-1553/60
SINBAD-FNS-DUCT. Dogleg Duct Streaming Experiment with 14 MeV Neutron Source (2004).

NEA-1553/61
SINBAD-FNS-OXYGEN. FNS/JAERI Time-of-Flight Experiment on Liquid Oxygen Slab With 14 MeV D-T Neutrons (1989).

NEA-1553/62
SINBAD-FNS-SKYSHINE. FNS/JAERI Measurement of Radiation Skyshine with D-T Neutron Source (March 2002 and March 2003).

NEA-1553/69
SINBAD-TUD-FNG-BS. TUD/FNG Measurement of Neutron and Photon Spectra in an ITER Bulk Shield Mock-up (1996).

NEA-1553/70
SINBAD-TUD-FNG-SIC. TUD/FNG measurement of neutron and photon flux spectra in a silicon carbide assembly (2001).

NEA-1553/71
SINBAD-FNG-HCPB. FNG Helium-cooled Pebble Bed (HCPB) Tritium Breeder Module (TBM) Mock-up (2005).

NEA-1553/72
SINBAD-FNS-C-CYLIND. Integral Experiment on a 60 cm-thick Graphite Cylindrical Assembly (FNS/JAERI clean benchmark) (1984).

NEA-1553/73
SINBAD-FNS-V. Neutron spectra and dosimetry, gamma-ray spectra and heating from a 25.4 cm cube of Vanadium irradiated with a D-T neutron source (FNS/JAERI clean benchmark) (1996).

NEA-1553/74
SINBAD-FNS-W. FNS/JAERI Clean Benchmark Experiment on Tungsten Cylindrical Assembly (1993).

NEA-1553/75
SINBAD-IPPE-FE. IPPE neutron transmission benchmark experiment with 14 MeV neutrons through iron shells.

NEA-1553/76
SINBAD-IPPE-V. IPPE neutron transmission benchmark experiment with 14 MeV neutrons through vanadium shells.

NEA-1553/77
SINBAD-OKTAVIAN/AL. Leakage Neutron and Gamma Spectra from Aluminium Sphere Pile With 14 MeV Neutrons (December 1988).

NEA-1553/78
SINBAD-OKTAVIAN-MN. Leakage Neutron and Gamma Spectra from 60 cm diameter Manganese Sphere Pile With 14 MeV Neutrons (July 1987).
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2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.
Program name Package id Status Status date
SINBAD FUSION-- NEA-1553/01 Tested 02-JUN-1996
SINBAD-ILL-FE NEA-1553/26 Arrived 01-DEC-2000
SINBAD-SB5-FUS NEA-1553/27 Arrived 01-DEC-2000
SINBAD-MEPHI NEA-1553/40 Tested 15-DEC-2005
SINBAD-KANT NEA-1553/41 Tested 13-MAR-2006
SINBAD-LI-BLANKET NEA-1553/43 Arrived 04-DEC-2008
SINBAD-FNS-BENCHM NEA-1553/45 Arrived 06-FEB-2009
SINBAD-TUD-FE NEA-1553/46 Tested 07-MAY-2010
SINBAD-TUD-FNG-W NEA-1553/47 Tested 05-MAY-2010
SINBAD-OKTAVIAN/NI NEA-1553/48 Arrived 03-MAY-2010
SINBAD-OKTAVIAN/FE NEA-1553/50 Tested 05-MAY-2010
SINBAD-OKTAVIAN/SI NEA-1553/52 Tested 05-MAY-2010
SINBAD-OKTAVIAN-W NEA-1553/53 Tested 07-MAY-2010
SINBAD-FNG-BLKT NEA-1553/54 Arrived 20-DEC-2011
SINBAD-FNG-DOSE-RATE NEA-1553/55 Arrived 20-DEC-2011
SINBAD-FNG-SIC NEA-1553/56 Arrived 20-DEC-2011
SINBAD-FNG-SS NEA-1553/57 Arrived 20-DEC-2011
SINBAD-FNG-STREAMING NEA-1553/58 Arrived 20-DEC-2011
SINBAD-FNG-W NEA-1553/59 Arrived 20-DEC-2011
SINBAD-FNS-DUCT NEA-1553/60 Arrived 20-DEC-2011
SINBAD-FNS-OXYGEN NEA-1553/61 Arrived 20-DEC-2011
SINBAD-FNS-SKYSHINE NEA-1553/62 Arrived 20-DEC-2011
SINBAD-TUD-FNG-BS NEA-1553/69 Arrived 21-DEC-2011
SINBAD-TUD-FNG-SIC NEA-1553/70 Arrived 21-DEC-2011
SINBAD-FNG-HCPB NEA-1553/71 Arrived 14-MAR-2012
SINBAD-FNS-C-CYLIND NEA-1553/72 Arrived 14-MAR-2012
SINBAD-FNS-V NEA-1553/73 Arrived 01-MAR-2012
SINBAD-FNS-W NEA-1553/74 Arrived 01-MAR-2012
SINBAD-IPPE-FE NEA-1553/75 Arrived 01-MAR-2012
SINBAD-IPPE-V NEA-1553/76 Arrived 01-MAR-2012
SINBAD-OKTAVIAN/AL NEA-1553/77 Arrived 01-MAR-2012
SINBAD-OKTAVIAN/MN NEA-1553/78 Arrived 01-MAR-2012

Machines used:

Package ID Orig. computer Test computer
NEA-1553/01 Many Computers Many Computers
NEA-1553/26 Many Computers
NEA-1553/27 Many Computers
NEA-1553/40 Many Computers Many Computers
NEA-1553/41 Many Computers Many Computers
NEA-1553/43 Many Computers
NEA-1553/45 Many Computers
NEA-1553/46 Many Computers Many Computers
NEA-1553/47 Many Computers Many Computers
NEA-1553/48 Many Computers Many Computers
NEA-1553/50 Many Computers Many Computers
NEA-1553/52 Many Computers Many Computers
NEA-1553/53 Many Computers Many Computers
NEA-1553/54 Many Computers
NEA-1553/55 Many Computers
NEA-1553/56 Many Computers
NEA-1553/57 Many Computers
NEA-1553/58 Many Computers
NEA-1553/59 Many Computers
NEA-1553/60 Many Computers
NEA-1553/61 Many Computers
NEA-1553/62 Many Computers
NEA-1553/69 Many Computers Many Computers
NEA-1553/70 Many Computers Many Computers
NEA-1553/71 Many Computers Many Computers
NEA-1553/72 Many Computers Many Computers
NEA-1553/73 Many Computers Many Computers
NEA-1553/74 Many Computers Many Computers
NEA-1553/75 Many Computers Many Computers
NEA-1553/76 Many Computers Many Computers
NEA-1553/77 Many Computers Many Computers
NEA-1553/78 Many Computers Many Computers
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3. DESCRIPTION

SINBAD is a new electronic database developed to store a variety of radiation shielding benchmark data so that users can easily retrieve and incorporate the data into their calculations. SINBAD is an excellent data source for users who require the quality assurance necessary in developing cross-section libraries or radiation trans- port codes. The future needs of the scientific community are best served by the electronic database format of SINBAD and its user- friendly interface, combined with its data accuracy and integrity. It has been designed to be able to include data from nuclear reactor shielding, fusion blankets and accelerator shielding experiments.
   
The guidelines developed by the Benchmark Problems Group of the American Nuclear Society Standards Committee (ANS-6) on formats for benchmark problem description have been followed by SINBAD. SINBAD data include benchmark information on (1) the experimental facility and the source; (2) the benchmark geometry and composition; and (3) the detection system, measured data, and an error analysis. A full reference section is included with the data. Relevant graphical information, such as experimental geometry or spectral data, is included. All information that is compiled for inclusion with SINBAD has been verified for accuracy and reviewed by two scientists.
NEA-1553/26
SINBAD-ILL-FE
=============
Purpose and Phenomena Tested
----------------------------
The purpose of this experiment was to compare measurements and calculations of fast-neutron leakage spectra from a spherical shell of iron to test the validity and accuracy of the neutron cross-section data.
  
Description of the Source and Experimental Configuration
--------------------------------------------------------
Two sources were used: (1) a californium-252 spontaneous fission source, and (2) a D-T fusion neutron source provided by a neutron generator. For the measurements using the D-T source, the flight tube of the neutron generator was inserted through a 9.5 cm diameter re entrant hole. For the measurements using the Cf-252 source, the reentrant hole was plugged with a steel cylinder, and the Cf-252 source was hung from a small steel holder attached to the plug. The holder positioned the Cf-252 source at the center of the iron sphere.
  
The iron sphere and the neutron detector were both situated 1 meter above the concrete floor with the detector positioned 200 cm from the center of the sphere.
  
The iron sphere contained 0.21% by weight of carbon and 0.47% by weight of manganese. The spherical shell of iron had an inner radius of 7.65 cm, an outer radius of 38.10 cm, and a number density of 0.0849 nuclei/barn-cm.

NEA-1553/27
SINBAD-SB5-FUS
==============
Purpose and Phenomena Tested
----------------------------
The experimental facility was built at Oak Ridge National Laboratory, building 6025, for the express purpose of conducting fusion reactor shielding experiments. The experiment contains shield configurations, which consisted of stainless steel or stainless steel and borated polyethylene slabs, was designed to measure neutron and gamma-ray spectra behind those shields due to 14-MeV neutrons incident on the shields. Provisions were made for measuring data for shields up to one meter thick, although the thickest measured was about 57 cm thick. The data obtained from the experiment could be used to test methods and data used in calculating neutron and gamma-ray transmission through fusion reactor shields having similar material compositions.
  
Description of the Source and Experimental Configuration
--------------------------------------------------------
The design of the facility and source was based upon a preconstruction computer analysis using the 2-Dimensional radiation transport code DOT. The room housing the experimental apparatus is roughly 7 m by 9 m lined with 92-cm-thick concrete personnel shields. In an adjacent room, an accelerator is connected via a drift tube to the target assembly for the neutron source generation. A large concrete test shield support structure, supports the test slab materials and reduces the background scattered radiation and was located toward the center of the room.
  
The source for the experiment was provided by a Sames Accelerator which directed 250-keV deuterons onto a 4 mg/cm2 titanium titride target. Neutrons with energies averaging about 14 MeV were produced with various energies and angles. A program used to calculate the source is shown in the full description of the SBE. The shield slabs are 152.4-cm square (Req=85.98 cm), but in the model they radially fill the 156.21-cm x 172.72-cm cavity (Req=92.67 cm). An additional thermal-neutron shield was placed behind the detectors to decrease the effect of neutron backscatter from and gamma-ray production in the concrete wall behind the detector. Six different shield configurations, housing various thicknesses and arrangements of stainless steel and/or borated polyethylene slabs, were tested with an additional configuration containing no test slabs ("no-shield" case). A total of sixty-four neutron and gamma-ray spectra were published. Data for three of the neutron measurements were thought unreliable but are included for completeness.

NEA-1553/40
SINBAD-MEPhI
============
Purpose and Phenomena Tested:
----------------------------
Measurements of neutron reaction rates, heating rates, neutron and photons spectra in iron shielding models with empty slits irradiated by 14-MeV neutrons.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The 14-MeV DT MEPhI neutron generator was the neutron source. The maximum neutron output equal 2E11 neutron per second by 200 keV accelerating potential. The following possible cases of a mutual position of two slits were considered during designing mock-ups with hollow slits in protective compositions:
  
1. The source neutrons can intersect two slits.
2. The neutrons of a source can intersect only one slit at minimum possible distance between the axis of the displaced slits.
3. The maximum position of the slits in the installation, when the path of source neutrons is less than a distance between slits.
  
In these cases at the 500 mm thickness of a mock-up and the 100 mm distance of a neutron source from a front surface the axis displacement of slits was 32.5, 70 and 120 mm. The models of shieldings were assembled from the steel milling blocks of the 25x200x500 mm3 and 50x200x500 mm3 sizes so that to avoid a direct run of neutrons on the touch surfaces of blocks. For this purpose the figured subsidiary blocks of a 1000 mm length, on which the upper part of a mock-up of a shield was installed, were made. The horizontal slit gap of a 20 mm size was made by erection of subsidiary blocks on two steel stops having the 20x25x500 mm3 sizes. The distance between the stops was selected from conditions of installation strength and their minimum influence to a condition of infinity of a slit size in a horizontal direction. For measurements of neutron and photon spectra, the horizontal channels of a 40 mm diameter each were arranged. The channels had a steel stub removing under insertion into the channel of a probe of a scintillation spectrometer. The shielding material is steel, having impurity content.

NEA-1553/41
SINBAD-KANT
===========
Purpose and Phenomena Tested:
----------------------------
The purpose at the time of the experiment was mainly to clarify the discrepancies found by other researchers in the effective neutron multiplication of bulk beryllium assemblies with central 14 MeV neutron sources. We recorded neutron leakage spectra from 5, 10 and 17 cm thick spherical beryllium shells from the source energy down to less than 10 eV. The experiment was performed at the Institut fuer Neutronenphysik und Reaktortechnik of Forschungszentrum Karlsruhe from 1990 to 1994.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The neutron source was the T(d,n)He-4 reaction. A titanium-tritium target on a copper backing was bombarded by a 150 keV deuteron beam. The source strength was monitored by recording associated alpha particles using a silicon surface barrier detector.
  
The spherical beryllium metal shells were placed in a concentric way around the source. The inner shell diameter was 10 cm and the outer up to 44 cm. All neutron leakage spectra were recorded at 60 degree emission angle with respect to the deuteron beam.

NEA-1553/43
SINBAD-LI-BLANKET
=================
Aim of the experiment:
---------------------
The neutronics experiments were performed to validate involved neutron cross sections, especially for tritium production.
  
Description of the Source and Experimental Configurations:
---------------------------------------------------------
A 14-MeV-fusion-neutron source (deuterium ions accelerated on to a T/Ti/Cu target) at Institut for Reactor Development (IRE) in Juelich was used to perform measurements of the tritium production with the aid of activation techniques and TLD detectors.
The space dependent neutron spectra, and energy deposition [3] are determined as well.
Several experimental configurations were tested. With/without inner Be neutron multiplication layer and outer Graphite reflector (Figure 7);
Li configuration - only Li blanket,
Be-Li configuration - Li blanket with Be inner layer,
Li-C configuration - Li blanket with outer graphite reflector,
Be-Li-C configuration - Li blanket with both Be and C layers.

NEA-1553/45
SINBAD-FNS-BENCHM
=================
Purpose and Phenomena Tested:
----------------------------
Collection of experimental data for fusion neutronics benchmark experiments performed at the Japan D-T neutron source facility FNS at JAERI.
  
The following experiments  are covered in the report: Tritium Breeding Ratio in Li, Pb-Li, Pb-Li-C, Be-Li, Be-Li-C spheres; leakage spectra from Be sphere and  Be-Li sphere; gamma energy spectra emitted from spheres with 14 MeV  neutron source; leakage neutron spectra from various sphere piles with 14 MeV neutrons; angular neutron flux spectra leaking from slabs; integral experiment on graphite cylindrical assembly; integral experiment on Li2O cylindrical assembly; integral experiment on Be cylindrical assembly.

NEA-1553/46
SINBAD-TUD-FE
=============
Purpose and Phenomena Tested:
----------------------------
Determination of spectral neutron flux, spectral photon flux and neutron time-of-arrival (TOA) flux penetrating and leaking iron slab assemblies (thickness: 30 cm; solid and with gap) irradiated with 14 MeV neutrons.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The neutron source was a 14 MeV d-T neutron generator operated in pulsed mode. The time distribution of the source neutrons was proportional to exp[-(t/1.4 ns)**2].
The Fe slabs had a front area of 100 cm x 100 cm and a thickness of 30 cm and were built up by bricks with dimensions of 20 cm x 10 cm x 5 cm.  
Three assemblies were built up:
A0 - no gap,  
A1 - vertical gap, distance: 10 cm from the centre, gap width: 5 cm, and  
A2 - vertical gap, distance: 20 cm from the centre, gap width: 5 cm.
The distance between neutron source and Fe slab was 19 cm. The distance between Fe slab and detector was 300 cm. The distance between neutron source and detector was 349 cm.
The angle between the d-beam of the neutron generator and an axis
crossing neutron source and centre of the slab was 74 degrees.
The detectors were positioned in a collimator shaped in such a way that all neutrons and photons leaking from the slab in direction of the detector could be observed.

NEA-1553/47
SINBAD-TUD-FNG-W
============
Purpose and phenomena tested:
-----------------------------
Transport data benchmark by determination of spectral neutron flux and spectral photon flux at four positions in a thick block of W irradiated with 14 MeV neutrons
  
Description of source and experimental configuration:
-----------------------------------------------------
The Frascati Neutron Generator [1] was used as 14 MeV D-T neutron source. The angular dependence of the source intensity is presented in Fig. 1. The angular dependence of the source energy distribution is given in Fig. 2. Note that these figures were obtained using the obsolete D-T source subroutine (source.for) and may vary slightly from those of the new recommended subroutine (DT_MCNP5.TXT for MCNP5, source.F and srcdx.F for MCNPX).
The geometry of the assembly is shown in Fig. 3. The angle between the deuteron beam of the neutron generator and an axis crossing neutron source and centre of the detector was 0 degrees. The dimensions of the W alloy block were 47 cm * 47 cm * 49 cm of length (z-axis). The tritium target of the neutron source was located at z = -5.3 cm. The W alloy of the central part had a density of 18.1 g/cm3 and an elemental composition of 95 wt-% of W, 1.6 wt-% of Fe and 3.4 wt-% of Ni.
  
Neutron and photon flux spectra were measured on the central axis of the assembly in four positions (P1,...,P4) at z = 5, 15, 25 and 35 cm.

NEA-1553/48
SINBAD-OKTAVIAN/NI
==================
Purpose and Phenomena Tested:
----------------------------
Neutron leakage spectra from a 32 cm diameter nickel sphere were measured between 30 keV and 15 MeV by the time-of-flight technique using a 14 MeV D-T neutron generator.
  
Description of the Source and Experimental Configuration:
--------------------------------------------------------
A 300 keV Cockcroft-Walton type accelerator, OKTAVIAN of Osaka University was used to accelerate deuterons to a kinetic energy of 245 keV. The deuteron beam was led through a narrow tube to the centre of a nickel sphere where pulsed 14.1 MeV monochromatic neutrons were produced by the T(d,n)He-4 fusion reaction. The source strength is angle dependent. The 32 cm diameter nickel sphere consisted of 99.63% Ni.

NEA-1553/50
SINBAD-OKTAVIAN/FE
==================
Purpose and Phenomena Tested:
----------------------------
Neutron leakage spectra from an iron sphere with a radius of 50.32 cm were measured by the time-of-flight technique using a 14 MeV neutron generator.
  
Description of the Source and Experimental Configuration:
--------------------------------------------------------
A Cockcroft-Walton type accelerator, OKTAVIAN of Osaka University was used to accelerate deuterons to a kinetic energy of 245 keV. The deuteron beam was led through a narrow tube to the centre of an iron sphere where pulsed 14.1 MeV monochromatic neutrons were produced by the d-t fusion reaction. The lower energy component of the source spectrum is about 5% of 14 MeV main component. Therefore the source is regarded to be 14 MeV monochromatic. The source strength is dependent on the angle. The 1 m diameter iron sphere consisted of carbon steel (98.69% Fe) plates.

NEA-1553/52
SINBAD-OKTAVIAN/SI
==================
Purpose and Phenomena Tested:
----------------------------
The leakage current spectrum from the outer surface of the sphere pile with 14 MeV neutrons normalized per source neutron was measured [1], [2], [3], [4]. The gamma-rays were produced from (n,xg) reactions.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The pulsed beam line of the intense 14 MeV neutron source facility OKTAVIAN [5] at Osaka University was used. Neutrons were produced by bombarding a 370 GBq tritium target with 250 keV deuteron beam. The energy spectrum of the neutron source was measured using the same detection system as for the leakage spectrum measurement. The spatial distribution of the emitted neutrons was measured for the target assembly, but for the purpose of the analysis an isotropic neutron source distribution is assumed.
  
The neutron spectra were measured with the time-of-flight (TOF) technique. A tritium neutron producing target was placed at the center of the pile. A cylindrical liquid organic scintillator NE-218 was used as a neutron detector, which was located about 11 m from the tritium target and at 55 deg. with respect to the deuteron beam axis. A pre-collimator made of polyethylene-iron multi-layers was set between the pile and the detector in order to reduce the background neutrons. The aperture size of this collimator was determined so that the whole surface of the pile facing the detector could be viewed.
  
Gamma-rays were detected with a cylindrical NaI crystal and the energy spectra were obtained from the unfolding process of the gamma-ray pulse-height spectra, using a response matrix of the NaI detector. The detector was located at 5.8 m distance from the neutron source and counted the gamma-rays emitted from the sphere. Time spectra of neutrons and gamma-rays from the sphere were measured simultaneously with the pulse-height spectra by means of a TOF technique.
  
The pile was made by filling a spherical vessels with grannular silicon. The stainless steel (JIS SUS-304) vessel with 61.0 cm outer diameter was equipped with a 20 cm inner diameter void at its center and a 11 cm diameter reentrant hole for the  target beam duct. The external and the internal vessel thickness was 0.5 cm and 0.2 cm respectively. The gamma-ray leakage spectrum has been also measured for a 40.0 cm diameter pile.
  
Granular silicon was at least 99.9% pure, with density of 1.29 g/cm3.
  
The assumed composition of stainless steel is 18.5 % Chromium, 70.4 % Iron and 11.1 % Nickel, with density of 7.86 g/cm3.

NEA-1553/53
SINBAD-OKTAVIAN-W
=================
Purpose and Phenomena Tested:
----------------------------
The leakage current spectrum from the outer surface of the sphere pile with 14 MeV neutrons normalized per source neutron was measured ([1], [2], [3], [4]). The gamma-rays were produced from (n,xg) reactions.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The pulsed beam line of the intense 14 MeV neutron source facility OKTAVIAN [5] at Osaka University was used. Neutrons were produced by bombarding a 370 GBq tritium target with 250 keV deuteron beam. The energy spectrum of the neutron source was measured using the same detection system as for the leakage spectrum measurement. The spatial distribution of the emitted neutrons was measured for the target assembly, but for the purpose of the analysis an isotropic neutron source distribution is assumed.
  
The neutron spectra were measured with the time-of-flight (TOF) technique. A tritium neutron producing target was placed at the center of the pile. A cylindrical liquid organic scintillator NE-218 was used as a neutron detector, which was located about 11 m from the tritium target and at 55 deg. with respect to the deuteron beam axis. A pre-collimator made of polyethylene-iron multi-layers was set between the pile and the detector in order to reduce the background neutrons. The aperture size of this collimator was determined so that the whole surface of the pile facing the detector could be viewed.
  
Gamma-rays were detected with a cylindrical NaI crystal and the energy spectra were obtained from the unfolding process of the gamma-ray pulse-height spectra, using a response matrix of the NaI detector. The detector was located at 5.8 m distance from the neutron source and counted the gamma-rays emitted from the sphere. Time spectra of neutrons and gamma-rays from the sphere were measured simultaneously with the pulse-height spectra by means of a TOF technique.
  
The pile was made by filling a spherical vessels with tungsten metal powder. The stainless steel (JIS SUS-304) vessel with 39.9 cm outer diameter was equipped with a 20 cm inner diameter void at its center and a 11 cm diameter reentrant hole for the  target beam duct. The vessel thickness was 0.2 cm everywhere.
  
Tungsten metal powder was at least 99.97% pure, with density of 4.43 g/cm3.
  
The assumed composition of stainless steel is 18.5 % chromium, 70.4 % iron and 11.1 % nickel, with density of 7.86 g/cm3.

NEA-1553/54
SINBAD-FNG-BLKT
===============
Purpose and Phenomena Tested
----------------------------
The neutronics experiments were performed to validate the International Thermonuclear Experimental Reactor (ITER) inboard shielding design.

Description of the Source and Experimental Configurations:
---------------------------------------------------------
Using a 14-MeV-fusion-neutron source (deuterium atoms accelerated onto a tritium target) at the ENEA Frascati Neutron Generator Facility was used to perform measurements of neutron penetration of a 94-cm-thick mockup of the first-wall, blanket, vacuum vessel, and toroidal field coils.  The block was made of copper, stainless steel/Perspex (polyethelyne-type material, C5O2H8) sandwich, followed by a smaller block made of alternating plates of copper and stainless steel simulating the magnet.

NEA-1553/55
SINBAD-FNG-DOSE-RATE
====================
Purpose and Phenomena Tested:
----------------------------
The purpose is to validate dose rate calculations for the International Thermonuclear Experimental Reactor (ITER). The experiment [1-4] was performed at the 14 MeV Frascati Neutron Generator (FNG) on a stainless steel/water assembly, in which a neutron spectrum was generated similar to that occurring in the ITER vacuum vessel. The mock-up was irradiated at FNG for sufficiently long time to create a level of activation which was, after shut down, followed by dosemeters for a cooling time assumed to be required for allowing personal access.

Description of Source and Experimental Configuration:
----------------------------------------------------
The 14-MeV d-T Frascati Neutron Generator (FNG, [5]) was the neutron source. The angular dependence of the source intensity is presented in Figure 1. The angular dependence of the source energy  distribution is shown in Figure 2. Note that these figures were obtained using the obsolete D-T source subroutine (source.for) and may vary slightly from those of the new recommended subroutine (DT_MCNP5.TXT for MCNP5, source.F and srcdx.F for MCNPX).

The x-y view of the geometry of the mock-up is outlined in Figure 3. It consists of a combination of slabs made from the water equivalent material Perspex and the stainless steel SS316 (simulating shield-blanket and vacuum vessel) and has a front cross-section area of 100 cm x 100 cm. The total thickness of the assembly is 71.83 cm. A cavity was arranged within the block, 119.8 mm (z) x 150 mm (x) x 126.0 mm (in the beam direction, y axis), behind a 22.37-cm-thick shield. A void channel (27.4 mm inner diameter) was included in front of the cavity to study the effect of streaming paths in the bulk shield (Fig.3). The channel wall was made of stainless steel AISI316 with 1.3 mm  thickness. A parallelepiped box was used to locate detectors inside the cavity, with 2-mm-thick bottom and lateral walls (stainless steel AISI316).

NEA-1553/56
SINBAD-FNG-SIC
==============
Purpose and Phenomena Tested:
----------------------------
The purpose is to validate the cross sections of Si and of C in the European Fusion File (EFF), as the SiC, in the form of ceramic matrix (SiC-fiber/SiC), is a candidate structural material for the fusion reactor and its development is pursued in the European Fusion Technology Program. The experiment [1] - [4] was performed at the 14 MeV Frascati Neutron Generator (FNG) on a monolithic, sintered SiC block.

Description of Source and Experimental Configuration:
----------------------------------------------------
The 14-MeV d-T Frascati Neutron Generator (FNG, [5]) was the neutron source. The angular dependence of the source intensity is presented in Table 1 and in Figure 1. The angular dependence of the source energy distribution is given in Table 2 and in Figure 2.  Note that these figures were obtained using the obsolete D-T source subroutine (source.for) and may vary slightly from those of the new recommended subroutine (DT_MCNP5.TXT for MCNP5, source.F and srcdx.F for MCNPX). The geometry of the mock-up is outlined in Figure 3. The experimental set-up consisted of a block of sintered SiC (45.72 cm x 45.72 cm, 71.12 cm in thickness), located in front of the FNG target, 5.3 cm from the neutron source. The SiC block was assembled with a total of 116 bricks. Inside the block, four experimental positions at different penetration depths were available to locate detectors of various types (activation foils, TLD holders, active spectrometers). The average weight density of the used SiC was 3.158 g/cm3, major impurities were boron (0.19 % in weight), aluminium (0.79 wt%) and iron (0.14 wt%).

NEA-1553/57
SINBAD-FNG-SS
=============
Purpose and Phenomena Tested:
----------------------------
The neutron transport in structural materials housing components of fusion devices are not well understood. Stainless steel is one such material, often used in the vacuum vessel and piping components. Neutron penetrations through large distances (60 cm) of stainless steel are reported.

Description of the Source and Experimental Configurations:
---------------------------------------------------------
A source of 14-Mev neutrons is generated by deuterons on a tritium target, for a total of 0.2789 neutrons per T(d,n) reaction, spread over a 60-degree forward angle, centered on the SS block face. The source is at 5.3 cm from the SS block face, 5 cm of air and 0.3 cm target source support structure.  The SS is of AISI 316 type, 70 cm deep, 100 cm wide x100 cm high.  SS plugs fill the central voids containing the detector foils.

NEA-1553/58
SINBAD-FNG-STREAMING
====================
Purpose and Phenomena Tested:
----------------------------
Determination of neutron reaction rates and of nuclear heating in a neutronic  mock-up of the International Thermonuclear Experimental Reactor (ITER) shielding system irradiated with 14-MeV neutrons, in presence of a streaming path.

Description of Source and Experimental Configuration:
----------------------------------------------------
The 14-MeV d-T Frascati Neutron Generator (FNG, [1]) was the neutron source. The angular dependence of the source intensity is presented in Figure 1. The angular dependence of the source energy distribution is shown in Figure 2.  Note that these figures were obtained using the obsolete D-T source subroutine (source.for) and may vary slightly from those of the new recommended subroutine (DT_MCNP5.TXT for MCNP5, source.F and srcdx.F for MCNPX).

The x-y view of the geometry of the mock-up is outlined in Figure 3. It consists of a combination of slabs made of the water equivalent material Perspex and the stainless steel AISI-316 (simulating shield-blanket and vacuum vessel) and has a front cross-section area of 100 cm x 100 cm. The total thickness of the assembly is 94.26 cm including a 1 cm thick Cu layer in front (simulating first wall). The assembly is provided with a channel with high aspect ratio (inner diameter a=28 mm, length l = 39.07 cm, channel wall thickness t = 1 mm stainless steel AISI316 ), (see Fig.3). At the end of the channel, a cavity is also realised. A parallelepiped stainless steel box fitting exactly the cavity is inserted to locate the detectors in their positions. The inner size of the box is 52 mm (z) x 148 mm (x) x 48 mm (in the deuterium beam direction-y). The cavity is symmetrically located with respect to the channel axis.

Behind this assembly a block of Cu and SS316 plates was arranged (simulating the coils for the toroidal magnetic field of the TOKAMAK; dimensions: depth 30.8 cm, area 47 cm x 47 cm). The rear part of the assembly was surrounded with a polythene shield covering also the last 30 cm of the Perspex/AISI316 block in order to reduce room-return background.

The following quantities are measured :

a - Measurements in the channel :
Neutron reaction rates by activation foils to monitor the flux gradient in the channel

b - Measurements in the cavity :
Neutron reaction rates by activation foils

c - Measurements behind the channel :
Neutron reaction rates by activation foils Nuclear heating

d - Measurements in the SC magnet : Nuclear heating

The detectors were placed on the axis of the d-beam of the neutron generator. All  measurements (a,b,c and d) were performed with the neutron source in axis with the channel/cavity, at 5.3 cm distance from the shielding block surface (ON-AXIS set-up). The activation foils measurements in the channel and in the cavity (a and b) were performed  also with the neutron source shifted with respect to the channel axis to simulate the effect of the extended neutron source from the plasma (OFF-AXIS set-up). The source lateral shift is 5.3 cm, i.e. equal to the distance between the target and  the mock-up surface. In this case the channel mouth is located at an angle of p/4 with respect to beam direction (see Fig.4).

NEA-1553/59
SINBAD-FNG-W
============
Purpose and Phenomena Tested:
----------------------------
The purpose is to validate the W cross sections in the European Fusion File (EFF), as tungsten is a candidate material for high flux component in the fusion reactor and its development is pursued in the European Fusion Technology Program. The experiment [1, 2] was performed at the 14 MeV Frascati Neutron Generator (FNG).

Description of Source and Experimental Configuration:
----------------------------------------------------
The 14-MeV d-T Frascati Neutron Generator (FNG, [3]) was the neutron source. The angular dependence of the source intensity is presented in Figure 1 and in Table 1. The angular dependence of the source energy distribution is shown in Figure 2 and is given in Table 2. Note that these figures were obtained using the obsolete D-T source subroutine (source.for) and may vary slightly from those of the new recommended subroutine (DT_MCNP5.TXT for MCNP5, source.F and srcdx.F for MCNPX). The geometry of the mock-up is outlined in Figure 4. It consisted of a block of a tungsten alloy, DENSIMET, produced by PLANSEE, in pieces of various shapes, assembled to obtain a size of about 42-47 cm (L) x 46.85 cm (H) and 49 cm in thickness and located in front of the FNG target, 5.3 cm from the neutron source. Most of the material (about 1.5 ton) is DENSIMET-176 type (92.3% W, 2.6% Fe, 4.2% Ni). A layer of DENSIMET-180 (about 0.25 ton, 7 cm height, composition 95.0% W, 1.6% Fe, 3.4% Ni) was used in the central part of the block were the measurements are done (Figs. 4-8), and contains the lateral access channels (diameter 5.2 cm) for locating detectors of the various types (activation foils, TLD holders, active spectrometers).

NEA-1553/60
SINBAD-FNS-DUCT
===============
Purpose and Phenomena Tested:
----------------------------
The experiment [1] was conducted at the FNS facility at JAERI to study the neutrons streaming through doubly bent ducts and estimate the uncertainties of calculations for the design of fusion reactors such as ITER, as well as to demonstrate the capability of the Monte Carlo transport calculations in the design of the fusion reactor shielding.

Description of the Source and Experimental Configuration:
--------------------------------------------------------
Two target rooms are available at FNS. The target that provides a neutron yield as large as about 4.0E12 n/s at full beam current was used in the experiment. The layout of the room is shown in Fig. 1 The experimental assembly was constructed in the wall separating the two target rooms. The assembly consists of an iron slab 1700 mm in height, 1400 mm in width, and 1800 mm in thickness. A doubly bent duct 300 mm x 300 mm in cross section was shaped through the assembly. The geometrical configuration of the dogleg duct streaming experiment is shown in Fig. 2. The first horizontal leg of the duct was set as high as the D-T neutron source. The second leg was connected vertically to the first with a right angle, and the third was horizontally to the second. The lengths of the legs were 1150 mm, 600 mm and 650 mm, respectively.

NEA-1553/61
SINBAD-FNS-OXYGEN
=================
Purpose and Phenomena Tested:
----------------------------
The angular neutron spectra leaking from the 200 mm slab of liquid Oxygen were measured at several angles.

Description of Source and Experimental Configuration:
----------------------------------------------------
The pulsed beam line of the intense 14 MeV neutron source facility at Fusion Neutronic Source (FNS) at JAERI was used. The D-T neutrons were produced by bombarding the tritium-titanium target with 350 keV deuteron beam. The energy spectrum of the neutron source is given in Table 2.

The slab assembly of liquid gas was set at the distance of 200 mm from the D-T neutron source. The assembly is made of a cylinder with a double wall for for thermal insulation. The void between the walls was evacuated. Aluminum foil layers between the walls were inserted to prevent the thermal radiation heating. The central area of the container walls was made of very thin windows of 300 and 465 mm in diameter to reduce neutron scattering for measuring angles. The thickness of the inner cylinder is 200 mm and the diameter 600 mm. The details of the container geometry are shown in Figure 1.

The atom density of Oxygen is 4.2947E-2 atoms/barn.cm.

NEA-1553/62
SINBAD-FNS-SKYSHINE
===================
Purpose and Phenomena Tested:
----------------------------
The D-T neutron skyshine experiments [1] have been carried out at the Fusion Neutronics Source (FNS) of JAERI. The radiation dose rate outside the target room was measured at distances up to 550 m from the D-T target point.

Description of Source and Experimental Configuration:
----------------------------------------------------
FNS is an accelerator based intense D-T neutron source. It has two target rooms: target room I has a fixed tritium target containing 3.7E+11 Bq tritium and the target room II has a rotating tritium target containing 3.7E+13 Bq tritium. The  target room I was used for the shyshine experiments. Figure 1 shows the cross-sectional view of target room I along a north-south cut. The target room is 15 x 15 m? wide and 9.1 m high. The thickness of the concrete of the roof and the vertical walls is 1.15 m and 2 m, respectively. The floor of the target room is an iron grating in order to reduce the neutron scattering from the floor. The tritium target is located 1.8 m above the floor, 5.5 m away from the south wall and 2.75 m from the west wall. The target room has a port in the roof with a size of 0.9 x 0.9 m? for the shyshine experiments.

This port is usually closed with a concrete plug. In the skyshine experiment, the concrete plug was removed and only 2 mm thick stainless steel plates closed the port in order to keep the reduced air pressure in the target room. The neutron attenuation by 2 mm stainless steel plates was negligible. FNS was operated with an acceleration voltage of 360 kV and a target current of 1.4 mA in this experiment. The source neutron yield was ~1.7E+11 n/s, which was monitored by the associated alpha particle measurement with a silicon surface barrier diode installed in the beam line.

Figure 2 shows the measurement points in the environment. FNS is located on a rather flat land only 150 m distance from the Pacific ocean. The north and east sides are surrounded with a forest of pine trees which are about 10 m tall. The neutron dose rate was measured at the distance from 20 to 550 m in the pine forest along the north direction, at distances from 20 to 140 m on a road along the south direction and at 200, 230 and 300 m
along the south west direction.

NEA-1553/69
SINBAD-TUD-FNG-BS
=================
Purpose and Phenomena Tested:
----------------------------
Determination of neutron and photon spectra in a neutronic mock-up of the International Thermonuclear Experimental Reactor (ITER) shielding system irradiated with 14-MeV neutrons.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The 14-MeV d-T Frascati Neutron Generator (FNG, [1]) was the neutron source.   
The geometry of the mock-up consists of a combination of slabs made from the water equivalent material Perspex and the stainless steel SS316 (simulating shield-blanket and vacuum vessel) and has a front cross-section area of 100 cm x 100 cm. The total thickness of the assembly is 94.26 cm including a 1 cm thick Cu layer in front (simulating first wall). Behind this assembly a block of Cu and SS316 plates was arranged (simulating the coils for the toroidal magnetic field of the TOKAMAK; dimensions: depth 30 cm, area 47 cm x 47 cm). The rear part of the assembly was surrounded with a polythene shield covering also the last 30 cm of the Perspex/SS316 block in order to reduce room-return background.
Neutron and photon spectra were determined in the mock-up on the central axis of the assembly at two positions:
  
Position A: Measurement behind the 6 cm thick Perspex layer inside a SS316 slab, at a total penetration depth 41.5 cm from the front of the assembly (Cu 1 cm, SS316 26.08 cm, and Perspex 14.42 cm).
  
Position B: Measurement in a SS316 layer at the total penetration depth 87.6 cm from the front of the assembly (Cu 1 cm, SS316 59.82 cm, and Perspex 26.78 cm).
  
The detectors were placed on the axis of the d-beam of the neutron generator.

NEA-1553/70
SINBAD-TUD-FNG-SIC
==============
Purpose and phenomena tested:
-----------------------------
Transport data benchmark by determination of spectral neutron flux and spectral photon flux at four positions in a thick block of SiC irradiated with 14 MeV neutrons.  
  
Description of source and experimental configuration:
-----------------------------------------------------
The Frascati Neutron Generator was used as 14 MeV D-T neutron source. The angular dependence of the source intensity and the angular dependence of the source energy distribution are given.
  
The geometry of the assembly is shown. The angle between the deuteron beam of the neutron generator and an axis crossing neutron source and centre of the detector was 0 degrees. The dimensions of the SiC block, composed of bricks, were 45.7 cm * 45.7 cm * 71.1 cm of length (x-axis). The tritium target of the neutron source was located at z = -5.3 cm. The SiC material was described in the calculations with a density of 3.158 g/cm3 and an elemental composition of 68.9 wt-% of Si, 30.8 wt-% of C, 0.19 wt-% of B, 0.079 wt-% of Al, and 0.014 wt-% of Fe.
  
The information on the concentration and heterogeneity of Boron may not be reliably. B-10 content strongly influences the thermal neutron flux and consequently the (n,gamma) reactions (e.g. gamma peak at 0.48 MeV from thermal neutron capture in B-10).
  
Neutron and photon flux spectra were measured on the central axis of the assembly at four positions (P1,...,P4) at x = 12.70, 27.94, 43.18 and 58.42 cm.

NEA-1553/71
SINBAD-FNG-HCPB
===============
Purpose and Phenomena Tested:
----------------------------
The scope of the experiment was the neutronics of the Helium-cooled Pebble Bed (HCPB) TBM mock-up for ITER.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The HCPB TBM mock-up [1, 2] was irradiated at the 14-MeV d-T Frascati Neutron Generator (FNG, [3]). The 14 MeV FNG neutron source was located 5.3 cm in front of the experimental block. The angular dependence of the source intensity is presented in Figure 1 and in Table 1. The angular dependence of the source energy distribution is ilustrated in Figure 2 and is given in Table 2. Note that these figures were obtained using the obsolete D-T source subroutine (source.for) and may vary slightly from those of the new recommended subroutine (DT_MCNP5.TXT for MCNP5, source.F and srcdx.F for MCNPX). Details of the FNG target are shown on Figure 3.
  
The geometry of the mock-up is outlined in Figure 4, Figure 5 and Figure 6. It consisted of an AISI 303 stainless steel (density = 7.954 g/cm3) box with external dimension 31.0 cm (x) x 29.0 cm (y) x 30.9 cm (z). The thickness of the steel box walls was 0.5 cm. The box was filled with metallic beryllium (density = 1.85 g/cm3) and contained two double layers made of breeder material (Li2CO3 powder - with natural Li, density = 1.123 g/cm3). The breeder layers had a thickness of 1.2 cm and were separated by 1 mm thick stainless steel walls.
  
The rear box was made of AISI-316 stainless steel with the 0.5 cm thick box walls, and the external dimensions of 31. cm (x) x 14.8 cm (y) x 30.9 cm (z). The box contains Li2CO3 powder (natural enrichment: 7.5% 6Li and 92.5% 7Li).  The total amount of Li2CO3 powder in the rear cassette is 11690.4 +/- 0.1 g, corresponding to a powder density of 0.9413 g/cm3.
  
The material compositions are given in Table 3. Note that the breeder material contains natural litium, i.e. with abundances 7.5 at.% 6Li, 92.5 at.% 7Li. For the ITER TBM design the use of Li4SiO4 or Li2TiO3 with enriched Li is considered.
  
The block was located on an aluminium support 5.3 cm in front of the FNG target. The box contained the lateral access channels for locating detectors of the various types (activation foils, Li2CO3 pellets).

NEA-1553/72
SINBAD-FNS-C-CYLIND
===================
Purpose and Phenomena Tested:
----------------------------
A cylindrical experimental assembly made of graphite was placed in front of D-T neutron source of Fusion Neutronic Source (FNS) facility at JAERI. Detectors were inserted in the assembly at several positions along the central axis of the cylinder. Nuclear responses related to both neutron and gamma-ray were measured with various experimental techniques. The measured quantities were: fission rate, reaction rate, neutron spectra and dose rate.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The mail sources of information used in this compilation were refs. [10], [2], [8] and [9].
Reactor-grade graphite blocks were stacked in thin wall (2 mm) aluminum tubes to form a cylindrical slab in the same manner as the lithium-oxide assembly [1]. The size of the graphite assembly was 31.4 cm in equivalent radius and 61.0 cm in thickness. The blocks used in this assembly were four types except near the central region where there was an experimental channel. The graphite blocks used were selected from the inventory so as to have the density with the deviation within +/- 2%. The average density was (1.641+/-0.015) g/cm3.
  
The experimental channel, a set of sheath and drawer, was made of the same grade graphite (100 x 100 x 1000 mm3 - 1.654+/-0.002 g/cm3) in order to save the changing time of detector position and for minimizing the personnel exposure for experimentalists. The experimental channel was placed at the central axis of the assembly. Therefore, this experimental assembly consisted of a single element, i.e., carbon, except the aluminum framework. Graphite blocks with experimental hole of 21 mm diam. Were made to allowinsertion of a detector.
  
The 80-degree beam line in the first target room of the FNS facility was used. A high speed water-cooled target [3] was set at the end of the beam line. A 7.4 x 1011 Bq (20Ci) Ti-T target was mounted on the target assembly. Neutrons were generated at the distance of 20 cm from the assembly surface on its central axis. The setting accuracy is estimated to be within +/-1 mm. The distances from the target to the west and south walls are 5.5 m, and those to the ceiling, the grating floor and the basement floor are 7.9, 1.8 and 3.8 m, respectively.
  
Neutron yields were determined by means of the associated alpha-particle detection method [4]. A small silicon surface-barrier detector with an aperture of about 1 mm diam. was mounted inside the beam line to detect the alpha-particle of 3T(d,n)4He reaction. Source characteristics, that is, neutron yield, angular distribution and spectra of the target assembly were measured by the time-of-flight technique [5], foil activation and a NE213 spectrometer [6].
  
A good agreement was obtained between neutron yield measured by different methods within the experimental error. An analysis by Monte Carlo calculation [7] also showed fairly good agreement with measured neutron energy spectra as well as angular distributions, the latter obtained by foil activation. Thus, the calculated source spectrum and the other characteristics were essentially confirmed and can be used as input information in the benchmark calculations. It should be noticed that the number of neutrons emitted toward 0 degree with respect to the d+ beam must be normalized as 1.1767 per unit D-T reaction at the target.

NEA-1553/73
SINBAD-FNS-V
============
Purpose and Phenomena Tested:
----------------------------
A 25.4 cm x 25.4 cm x 25.4 cm cube vanadium assembly was irradiated in the D-T neutron source of Fusion Neutronic Source (FNS) facility at JAERI. Neutron spectra down to 1 eV, dosimetry reaction rates, gamma-ray spectra and gamma- ray heating rates were measured at 3 positions up to 17.78 cm inside the vanadium assembly.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The experimental configuration was basically the same as those used for some previous clean benchmark experiments like the Lithium-Oxide, Graphite, Copper, Beryllium, Tungsten assemblies.
The vanadium assembly was a cube with a side of 254 mm. Four side surfaces and a rear surface of the vanadium assembly was covered with a graphite reflector with the thickness of 50.8 mm to reduce leakage neutrons from the rather small assembly and incoming background neutrons from the outside. Purity of vanadium was higher than 99.7 %.
  
The experimental assembly was located in front of the neutron source at a distance of 200 mm from the target. The tritium-titanium target of ~3.7E11 Bq was bombarded by a deuteron beam of 350 keV energy to produce D-T neutrons. The number of source D-T neutrons generated during each measurement was determined by the alpha-particle detector with an accuracy of 2~3 %.
  
The D-T neutron source can be roughly described as an isotropic point 14-MeV neutron source.  The source neutron spectrum and intensity, however, depend slightly on the emission angle. The angle-dependent source characteristics were investigated experimentally and theoretically in detail, and a source subroutine for the Monte Carlo transport code MCNP-4 has been prepared to simulate the source condition precisely. This routine is listed in [3]. A comparison between the measured and calculated angular distribution is given in Fig. 3.2.7. in [4]. As an adequate alternative to the use of the MCNP routine it is suggested that the source spectrum of neutrons emitted toward the 0 degree direction with respect to the deuteron beam direction can be used for the incident neutron spectrum on the whole front surface of the assembly. No quantitative estimation of difference between both sources could be though found in the available literature. The si1 and sp1 cards indicate upper neutron energies of the energy bins in MeV and probabilities in the bins, respectively. The dir and vec parameters with the sb2 card are used for variance reduction with the source biasing method. The weight of a source neutron specified by the wgt parameter, 1.1261, is larger than 1.0 because more D-T neutrons are emitted to the forward direction than in the backward direction with respect to the deuteron beam direction.  
  
The tritium target region is also a source of gamma-rays, namely, target gamma- rays, created by interaction of the source neutrons with structural materials of the target. Consideration of the target gamma-rays, however, is not needed in calculations of gamma-ray heating rates because contribution of the target gamma-ray to the measured heating rates has been already subtracted in the experimental data.  On the other hand, the target gamma-ray contribution is involved in the measured gamma-ray spectra. The contribution at the detector positions deep inside the experimental assemblies is negligible, representing at most few percents, because the experimental assembly largely attenuates the target gamma-rays. Accordingly, it is not necessary to consider the target gamma-rays in transport calculations for the clean benchmark experiments.
   
Since the V assemblies were located at least 4 m from the experimental room walls and floor, the contribution of the background neutrons and gamma-rays coming from the room walls and floor on the measured quantities was negligibly small.

NEA-1553/74
SINBAD-FNS-W
============
Purpose and Phenomena Tested:
----------------------------
A tungsten cylindrical assembly (diameter = 629 mm, height = 507 mm) was irradiated in the D-T neutron source of Fusion Neutronic Source (FNS) facility at JAERI. Neutron spectra down to 5 keV, dosimetry reaction rates, gamma-ray spectra and gamma-ray heating rates were measured at 3 positions up to 380 mm inside the tungsten assembly.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The experimental configuration was basically the same as those used for some previous clean benchmark experiments like the Lithium-Oxide, Graphite, Copper, Beryllium, Vanadium assemblies. The Tungsten assembly was made by stacking bricks of (50.7-50.8 mm) with thin aluminium support frame in quasi-cylindrical shape (diameter = 629 mm, height = 507 mm). The tungsten is not pure, but an alloy with a small amount of nickel and copper.
  
The experimental assembly was located in front of the neutron source at a distance of 200 mm from the target. The tritium-titanium target of ~3.7E11 Bq was bombarded by a deuteron beam of 350 keV energy to produce D-T neutrons. The number of source D-T neutrons generated during each measurement was determined by the alpha-particle detector with an accuracy of 2~3 %.
  
The D-T neutron source can be roughly described as an isotropic point 14-MeV neutron source.  The source neutron spectrum and intensity, however, depend slightly on the emission angle. The angle-dependent source characteristics were investigated experimentally and theoretically in detail, and a source subroutine for the Monte Carlo transport code MCNP-4 has been prepared to simulate the source condition precisely. This routine is listed in [3]. A comparison between the measured and calculated angular distribution is given in Fig. 3.2.7. in [4]. As an adequate alternative to the use of the MCNP routine it is suggested that the source spectrum of neutrons emitted toward the 0 degree direction with respect to the deuteron beam direction can be used for the incident neutron spectrum on the whole front surface of the assembly. No quantitative estimation of difference between both sources could be though found in the available literature. The 0 degree source neutron spectrum is shown, and input cards of MCNP for description of the source spectrum are shown in the MCNP sample input file. The si1 and sp1 cards indicate upper neutron energies of the energy bins in MeV and probabilities in the bins, respectively. The dir and vec parameters with the sb2 card are used for variance reduction with the source biasing method. The weight of a source neutron specified by the wgt parameter, 1.1261, is larger than 1.0 because more D-T neutrons are emitted to the forward direction than in the backward direction with respect to the deuteron beam direction.  
  
The tritium target region is also a source of gamma-rays, namely, target gamma-rays, created by interaction of the source neutrons with structural materials of the target. Consideration of the target gamma-rays, however, is not needed in calculations of gamma-ray heating rates because contribution of the target gamma-ray to the measured heating rates has been already subtracted in the experimental data.  On the other hand, the target gamma-ray contribution is involved in the measured gamma-ray spectra. The contribution at the detector positions deep inside the experimental assemblies is negligible, representing at most few percents, because the experimental assembly largely attenuates the target gamma-rays. Accordingly, it is not necessary to consider the target gamma-rays in transport calculations for the clean benchmark experiments.
  
Since the W assemblies were located atleast 4 m from the experimental room walls and floor, the contribution of the background neutrons and gamma-rays coming from the room walls and floor on the measured quantities was negligibly small.

NEA-1553/75
SINBAD-IPPE-FE
==============
Purpose and Phenomena Tested:
----------------------------
Neutron leakage spectra between 50 keV and 15 MeV from five iron shells were measured by the time-of-flight technique using a 14 MeV neutron generator. The spheres had radius from 4.5 to 30.0 cm and wall thicknesses from 2.5 to 28.0 cm.
The experiments were performed in period 1989 to 1995.
  
Description of Source and Experimental Configuration
----------------------------------------------------
A Cockroft-Walton type accelerator, the KG-0.3 pulse neutron generator in Obninsk, was used to accelerate deuterons to a maximum kinetic energy of 280 keV.
The deuterons were led through a conical aluminum tube of only 0.5 mm wall thickness and collimated by a diaphragm with an 8 mm hole to a solid Titanium-Tritium target based on a copper radiator 0.8 mm thick and with diameter 11 mm. Beam spot diameter was 5 mm.
The center of the target is located in the geometrical center of the iron shell. For monitoring the neutron source strength, the alfa particles   generated in the deuterium-tritium reaction were detected at 175 deg. Through a 1 mm diameter collimator by a silicon surface barrier (SSB) detector.
The ion pulse width is 2.5 ns. The repetition period of the pulse can be set arbitrarily to multiples of 200 ns. The mean beam current for 800 ns. Period is 1 microampere.
The mean energy and yield of the '14 MeV' neutron source peak are slightly angle dependent.
Five spherical shells of pure iron were available for the experiment.
The weight of every sphere was measured and atom density was calculated.

NEA-1553/76
SINBAD-IPPE-V
=============
Purpose and Phenomena Tested
----------------------------
Neutron leakage spectra between 50 keV and 15 MeV from two vanadium shells were measured by the time-of-flight technique using a 14 MeV neutron generator. The spheres had radius of 5 and 12 cm and wall thicknesses of 3.5 and 10.5 cm. The experiments were performed in two phases in 1996 and 1997.
  
Description of Source and Experimental Configuration
----------------------------------------------------
A Cockroft-Walton type accelerator, the KG-0.3 pulse neutron generator in Obninsk, was used to accelerate deuterons to a maximum kinetic energy of 310 keV.
The deuterons were led through a conical aluminum tube of only 0.5 mm wall thickness and collimated by a diaphragm with an 8 mm hole to a solid Titanium-Tritium target based on a copper radiator 0.8 mm thick and with diameter 11 mm. Beam spot diameter was 5 mm.
The center of the target is located in the geometrical center of the vanadium shell. For monitoring the neutron source strength, the alpha particles generated in the deuterium-tritium reaction were detected at 175 deg. Through a 1 mm diameter collimator by a silicon surface barrier (SSB) detector.
The ion pulse width is 2.5 ns. The repetition period of the pulse can be set arbitrarily to multiples of 200 ns. The mean beam current for 800 ns. Period is 1 microampere.
The mean energy and yield of the '14 MeV' neutron source peak are slightly angle dependent.
Two spherical shells of high purity vanadium were available for the experiment.
The density of every sphere was measured and found to agree with the literature value of 6.09 g/cm3.

NEA-1553/77
SINBAD-OKTAVIAN/AL
==================
Purpose and Phenomena Tested:
----------------------------
The leakage current spectrum from the outer surface of the sphere pile with 14 MeV neutrons normalized per source neutron was measured. The gamma-rays were produced from (n,xgamma) reactions.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The pulsed beam line of the intense 14 MeV neutron source facility OKTAVIAN [3] at Osaka University was used. Neutrons were produced by bombarding a 370 GBq tritium target with 250 keV deuteron beam. (Note: 243 keV is stated in [5] for the photon spectrum measurement, but the same neutron source spectrum is specified for the analysis). The energy spectrum of the neutron source was measured using the same detection system as for the leakage spectrum measurement. The spatial distribution of the emitted neutrons was measured for the target assembly, but for the purpose of the analysis an isotropic neutron source distribution is assumed.
  
The neutron spectra were measured with the time-of-flight (TOF) technique. A tritium neutron producing target was placed at the center of the pile. A cylindrical liquid organic scintillator NE-218 was used as a neutron detector, which was located about 11 m from the tritium target and at 55 deg. with respect to the deuteron beam axis. A pre-collimator made of polyethylene-iron multi-layers was set between the pile and the detector in order to reduce the background neutrons. The aperture size of this collimator was determined so that the whole surface of the pile facing the detector could be viewed.
  
Gamma-rays were detected with a cylindrical NaI crystal and the energy spectra were obtained from the unfolding process of the gamma-ray pulse-height spectra, using a response matrix of the NaI detector. The detector was located at 5.8 m distance from the neutron source and counted the gamma-rays emitted from the sphere. Time spectra of neutrons and gamma-rays from the sphere were measured simultaneously with the pulse-height spectra by means of a TOF technique.
  
The pile was made by filling a spherical vessels with aluminium powder. The stainless steel (JIS SUS-304) vessel with 39.9 cm outer diameter was equipped with a 20 cm inner diameter void at its center and a 11 cm diameter reentrant hole for the  target beam duct. The vessel thickness was 0.2 cm everywhere.
  
Aluminium powder was at least 99.7% pure with impurities consisting of less than 0.2% iron, less than 0.15% silicon and less than 0.01% copper.
  
The assumed composition of stainless steel is 18.5 % Chromium, 70.4 % Iron and 11.1 % Nickel.

NEA-1553/78
SINBAD-OKTAVIAN-MN
=================
Purpose and Phenomena Tested:
----------------------------
The neutron and gamma leakage spectra from the outer surface of the sphere pile with 14 MeV neutrons normalized per source neutron was measured ([1], [2]). The gamma-rays were produced from (n,xg) reactions.
  
Description of Source and Experimental Configuration:
----------------------------------------------------
The pulsed beam line of the intense 14 MeV neutron source facility OKTAVIAN [1] at Osaka University was used. Neutrons were produced by bombarding a 370 GBq tritium target with 250 keV deuteron beam. The energy spectrum of the neutron source was measured using the same detection system as for the leakage spectrum measurement. The spatial distribution of the emitted neutrons was measured for the target assembly. The neutron source spectrum is given in Table 1. More information about the source neutrons emission is given in [3] and [4]. The neutron spectra were measured with the time-of-flight (TOF) technique. A tritium neutron producing target was placed at the center of the pile. A cylindrical liquid organic scintillator NE-218 was used as a neutron detector, which was located about 11 m from the tritium target and at 55 deg. with respect to the deuteron beam axis. A pre-collimator made of polyethylene-iron multi-layers was set between the pile and the detector in order to reduce the background neutrons. The aperture size of this collimator was determined so that the whole surface of the pile facing the detector could be viewed. The details of the experimental set-up are shown in Figure 1. A better representation of the collimator system is available in [4]. Gamma-rays were detected with a cylindrical NaI crystal and the energy spectra were obtained from the unfolding process of the gamma-ray pulse-height spectra, using a response matrix of the NaI detector. The detector was located at 5.8 m distance from the neutron source and counted the gamma-rays emitted from the sphere.    The energy spectrum of gamma-rays at the source is shown in Table 2. Time spectra of neutrons and gamma-rays from the sphere were measured simultaneously with the pulse-height spectra by means of a TOF technique. The pile was made by filling a spherical vessels with manganese metal powder. The vessel was made of soft stell (JIS SS-41). Inner diameter and wall thickness of the vessel are 60 cm and 0.5 cm respectively. A reentrant hole for a beam duct is equipped, the diameter of which is 5.1 cm up to the centre of the vessel. The details of the pile geometry are available in Figure 2. Manganese metal powder was at least 99.95% pure, with density of 4.37 g/cm3.
top ]
4. METHODS
NEA-1553/26
SINBAD-ILL-FE
=============
Measurement System and Uncertainties
------------------------------------
The neutron spectrum measurements were made with a 5 cm x 5 cm glass-encapsulated NE-213 scintillator. The proton-recoil spectrometry system was chosen because it does not require an elaborate pulsed neutron source.
  
The total estimated systematic error in the normalization of the measurements is 8% standard deviation for a Cf-252 source within the iron spherical shell.
  
Description of Results and Analysis
-----------------------------------
The energy range between 1.0 and 15 MeV was covered by the NE-213 scintillator. Background contributions to the recoil spectra were measured by placing a paraffin shadow cone midway between the detector and the spherical assembly.
  
The spectra were unfolded by the FORIST computer code, which is a modified version of the COOLC and FERDoR computer codes. These modifications resulted in unfolded spectra having optimized energy resolution and increased accuracy through the use of an iterative smoothing technique.
  
The ANISN one-dimensional discrete ordinates neutron transport code was used to calculate the leakage spectra.

NEA-1553/27
SINBAD-SB5-FUS
==============
Measurement System and Uncertainties
------------------------------------
Measurements were made with an NE-213 detector manufactured at the Oak Ridge National Laboratory. Neutron and gamma-ray responses were distinguished through pulse-shape discrimination. The measured data were unfolded with the FERD code, and results were reported as point data on a fine energy grid and for a two standard deviation band about the mean values. The uncertainties are not necessarily the uncertainties of the measurements. Quoted uncertainties for the maximum shield thickness examined include: NE-213 efficiency (5%), alpha calibration (3%), neutron detector position (1%), nine stainless steel slabs (3%), and two borated polyethylene slabs (1%).
  
Description of Results and Analysis
-----------------------------------
The various measurements are described using an x-y-z coordinate location of the detector position, the configuration of material slabs and whether it is a gamma or neutron measurement. Data for runs R148NS.008, C171NS.016, and C171GS.012 are considered unreliable, as repeated measurements were substantially different and more consistent with other results. The normalized spectra (MeV-1 cm-2) per source neutron and energy structure are tabulated and plotted for a total of 64 measurements, behind 7 configurations, at up to 10 different angles or lateral (z positions).

NEA-1553/40
SINBAD-MEPhI
============
Measurement System:
-------------------
The following nuclear reactions were chosen for the measurements: In-115(n,n'), Zn-64(n,p), Al-27(n,alpha), Fe-56(n,p), F-19(n,2n) and U-233(n,f). In defining detector mass and sizes the following peculiarities of measurements in slit shield compositions were taken into account:
- considerable neutron field gradients near slit boundaries perpendicular to their directions,
- increase of reaction rate gradients near the slits with their energy threshold increase,
- insignificant neutron field gradients along the direction of the slits,
- decrease of reaction rates by a factor of 2-3 with a high-energy threshold in moving away from the slit edge.
  
Detector masses necessary for reaction rates measurements with a statistical error in the range of (3-5)% as well as their form and sizes needed for obtaining necessary spatial resolution of the measured quantities were determined on the basis of the calculations. To reduce the number of detectors, modules in the form of a parallelepiped with edge sizes equal to 2x5x30 mm3 were manufactured. In the experiments detectors were put at the back surface of the shielding compositions. The error of coordinate setting of the detector did not exceed 0.2 mm. The detectors length is equal to 60 mm, their height being 5 mm. The detectors were arranged at the back surface of the assembly in such a way that 60x5 mm2 faces were parallel to the slit surface. The measurements along X and Y axis were made.
  
In order to measure heating rate in an iron composition material a technique based on application of thermoluminescent detectors was used. Harshaw-2080 which included a heating device, a photoelectric detector and necessary electronics circuits was used to process TLD.
  
The heating device provides uniform volume heating with a constant linear velocity in the temperature range of (50-400)deg.C. Detectors were some powder poured into a glass ample with inner and outside diameters equal to 1 mm and 2 mm correspondingly, the height being equal to (15-20) mm. Contribution of a gamma-radiation, arising in a target unit, was accounted in heating rates for a mock-up of a uniform iron shield.
  
Two stilbene crystals of cylindrical shape with the sizes of 30x30 mm2 and 40x40 mm2 are used in the basic version of a spectrometer for measurements in bulk composition. For measurements on the back surface the stilbene crystal of the size 50x40 mm3 was used.
  
All measured neutron and photon quantities were normalized to one neutron source and one nucleus (or gramme).
  
Description of Results and Analysis:
-----------------------------------
Calculation and experimental studies showed that there was some dependence of nuclear reaction rates on the size of an ion spot which defined the neutron source. Mostly this was observed for reaction rates with high energy thresholds. To avoid this effect high voltage stabilization at a level less than 0.5% at the neutron generator in MEPhI was required. It is a complicated technical problem this is the reason why all the slits in target structures were located horizontally and a diaphragm was introduced into the target block. It was proved experimentally that in case of 500 mm thick shield and 20 mm transverse slit the largest width of a square diaphragm limiting falling ion flux down the target was equal to 13.8 mm. Thus there is a compromise between the size of a neutron source and its yield decrease.
  
The measurement of an absolute neutron yield is based on counting of alpha particles accompanying neutrons when the latter are produced. Alpha particle radiometers are placed in the ion beam conduit and have two independent channels with different diaphragm diameters. 0.3 mm thick scintillation crystals CsJ(Tl) are placed behind these diaphragms. The accuracy of the neutron output corresponds to 2.5% for one sigma. The nuclear reaction rate measurement system includes eight gamma-spectrometric tracts on NaJ(Tl) crystals with 63x63 mm2 sizes that are connected with an amplitude analyzer plate entered into a personal computer.
  
Heating rate in shield compositions is determined by an absorbed dose of ionizing radiations in its material components. In the present experiments the shield material is steel, the main component of which is iron. The method of measuring the absorbed dose should meet the following demands:
- high sensitivity;
- small distortion of neutron and gamma-fields;
- high spatial resolution;
- wide linear range of absorbed dose measurements;
- error in measuring absolute value of the absorbed dose should be at (8-10)% level.
  
At present TLD irradiation in shield mock-ups is considered to be the most suitable for these purposes. Bragg-Grey relations are taken to be a methodical base because they correlate with the dose absorbed in the detector and its environment. TLD used in the present experiments basic characteristics makes it possible to obtain values of the absorbed dose from gamma-quanta in the material of shield compositions by an effective atomic number by means of interpolation. It should be pointed out that light output of TLD which is being measured after irradiating the latter in shield compositions, will correspond to the absorbed dose of not only gamma but also neutron radiation.
  
For MCNP-4c2 calculation the MCNP-4c2 input files are provided. The co-ordinate axis in calculation and experiment have not mutual correlation.

NEA-1553/41
SINBAD-KANT
===========
Measurement System and Uncertainties:
-------------------------------------
Various detectors were employed to cover, in separate runs, the wide energy range: an NE-213 liquid scintillator detector for energies above 3 MeV, several different proton recoil proportional counter tubes for the range from 6 MeV down to about 50 keV, and the time-of-flight (TOF) technique for energies below 100 keV. The background of wall scattered etc. neutrons was always measured separately using a shadow cone.
  
The 1-standard-deviation uncertainty of every value in the spectra, including all systematic and statistical sources of uncertainties, is estimated at 8%.
  
In addition to the above spectrum measurements, the spectra were also recorded using the carefully calibrated Bonner sphere spectrometer of Physikalisch-Technische Bundesanstalt, Braunschweig (PTB). This technique offers only a modest energy resolution, but is sure to observe any neutrons between the source energy and thermal energy. It was employed here mainly to obtain an independent result on the total neutron multiplication. They agreed satisfactorily with those of the more detailed spectral measurements.
  
Description of Results and Analysis:
-----------------------------------
The results, in the form of tabulated 178-group spectra, are found in files fzkbe.tb1, fzkbe.tb2 and fzkbe.tb3). Some more details of the experiment are found in file fzk-be_e.htm. For analyses, an MCNP model of the experiment with 17 cm shell thickness is given in file kantmcnp.i. It can easily be modified for the 5 or 10 cm shell by changing the outer shell diameter accordingly.
  
Our own analysis [M95] revealed that the effective neutron multiplication generally agreed well with 3-dimensional calculations using recent (at the time) evaluated nuclear data. The relatively old Young and Stewart beryllium evaluation[Y79], used in the EFF-1 library, appeared to underestimate the emission of

NEA-1553/43
SINBAD-LI-BLANKET
=================
Measurement Systems and Uncertainties:
-------------------------------------
Li2CO3  samples, TLD detectors and activation foils were placed along the central perpendicular to the model axis experimental channel.
Several irradiations sessions were performed in the range 1014 to 1016  14 MeV source neutrons.
Tritium activity was measured by liquid scintillation method [6], gas counting method [5] and the use of LiF TLD detectors [3], [4].
Foils activity were recorded from calibrated Ge(Li) detectors.
Errors from counting statistics, uncertainty of detector calibrations, are negligible comparing with the source intensity uncertainty(15-20%).
A set of TLD-100/600/700 detectors were used for measuring the nuclear heating as a function of radius inside the blanket. These were calibrated by a 137Cs source.
For the liquid scintillation method, foil activation and TLD measurements neutron source intensity was measured by the activation method with the uncertainty of the 58Ni(n,2n)57Ni and 19F(n,2n)18F cross sections of 15-20%.
For gas counting method 23Na(n,2n)22Na and 89Y(n,2n)88Y reactions were used. Uncertainty in these cross sections are bigger - 20-30%.
  
Description of Results and Analysis:
-----------------------------------
Li2CO3 samples in Al sample holders, TLD detectors and activation foils were placed at different distances from the model axis(10 to 60 cm, to 73 cm in graphite reflector).
Results are given with the random error. Systematic errors are due to the uncertainty of the activation cross sections used for neutron source intensity measurements. MORSE code was used to compute tritium production rates in Li.
As original Morse input is not available, recently a simplified MCNP inputs for each configuration (Li,  Be-Li, Be-Li-C,Li-C ),are constructed (Figure 7) and some of the results are here presented.
The experimental results of the tritium production, and saturation activities are compared with MCNP and FISPACT calculations.

NEA-1553/46
SINBAD-TUD-FE
=============
Measurement System:
------------------
A NE213 scintillator was employed for simultaneously measuring the spectral neutron flux, the spectral photon flux, and the neutron time-of-arrival spectrum for neutron energies of E > 1 MeV and photon energies of E > 0.2 MeV. For each registered event the pulse-height, the time-of-arrival, and a pulse-shape parameter were recorded to distinguish between neutrons and photons. [1-5]
Pulse-height distributions from three different hydrogen-filled proportional detectors and a stilbene scintillator were used for determining the neutron flux spectra for energies ranging from 30 keV up to about 2.3 MeV, overlapping with the NE213 spectra.
  
Description of Results and Analysis:
-----------------------------------
Neutron energy spectrum:
The NE213 pulse-height spectra were unfolded by the DIFBAS code [6] with a response matrix developed at Physikalisch-Technische Bundesanstalt Braunschweig [7], to obtain spectral fluxes. The evaluation procedure of the proton recoil spectra from the proportional detectors and from the stilbene scintillator consisted in an iterative differentiation with inclusion of corrections (wall effect, non-linear light output function, anisotropy effect of stilbene, energy dependent sensitivity of the stilbene crystal, correction of neutron reactions on carbon resulting in alpha-particles).
  
Neutron time-of-arrival spectrum:
The neutron time-of-arrival distributions after the start pulse of the 14 MeV source neutrons (t=0), recorded by the NE213 scintillation detector, are presented vs. calibrated time scale. They are neither evaluated (e.g. with the detector efficiency nor corrected.
  
Photon energy spectrum:
Photons produced by neutrons in the Fe assembly arrive at the detector in the time range between 10 ns and 50 ns; whereas those photons produced by neutrons in the walls and floor of the experimental hall, in the detector collimator, and in the detector itself, arrive later. Therefore, the pulse-height distribution from the NE213 detector for photons was taken only in this time-window and was unfolded with the DIFBAS code [6] with a response matrix from Physikalisch-Technische Bundesanstalt Braunschweig [8].

NEA-1553/47
SINBAD-TUD-FNG-W
============
Measurement system:
------------------
Neutron and photon pulse-height spectra were measured simultaneously using an NE 213 scintillation spectrometer. The dimensions of the cylindrical active volume of the detector were 3.8 cm in both height and diameter. Its material had a mass density of 0.874 g/cm3 and an elemental composition of 54.8 at-% H and 45.2 at-% C. The scintillator was coupled to a photomultiplier by means of a 50 cm long light guide. When the detector was located at one of the positions (P 2 in Fig. 3), the other ones were filled with rods of W alloy.
  
Description of results and analysis:
-----------------------------------
Data evaluation: The DIFBAS code developed at PTB Braunschweig [2,3] was employed for unfolding the measured pulse-height distributions in order to generate the neutron and photon flux spectra. They were obtained as absolute fluxes, as the response matrices have been determined on an absolute scale by detailed simulations of experimental distributions from mono- energetic neutron and photon sources with Monte Carlo codes.
  
Calculations:
A computational analysis was performed with the Monte Carlo code MCNP- 4C [4] using a full 3D model of the assembly, the neutron generator and the experimental hall. Nuclear data were taken from the FENDL/MC-2.0 data library [5]. The spectra were calculated as average flux in the scintillator volume by means of the track length estimator of MCNP.
Results are shown in Figs. 4, 5, 6, 7, 8, 9, 10 and 11. More details are given in Ref. [6].
For the activation foil and TLD measurements on the same Tungsten block see FNG Tungsten Experiment.

NEA-1553/48
SINBAD-OKTAVIAN/NI
==================
Measurement System and Uncertainties:
------------------------------------
The detectors used were:
  
Detector       Diameter Thickness   Energy covered
                 (cm)     (cm)         (MeV)  
-------------------------------------------------------------------
NE-213 scint.   12.7      5.08       0.03-15

NE-213 scint.   5.08      5.08        (14)   neutron source monitor
-------------------------------------------------------------------
Measurements of neutron spectra were carried out for high energy region (between 1 MeV and 15 MeV) and low energy region (30 keV to 15 MeV). Energy dependent statistical errors, large below about 0.04 MeV, elsewhere 1 - 10%. The uncertainty of the source spectrum in the 14 to 15 MeV energy range is about 8 %.
  
Description of Results and Analysis:
-----------------------------------
Leakage current energy spectrum of neutrons was measured in "absolute values" by the time-of-flight technique between 30 keV and 15 MeV, about 9.5 m from the sphere centre. The detectors could view the entire surface of the sphere.
  
Run-to-run neutron source monitoring was done using NE-213 detector which viewed directly the d-t source and was located at 482 cm from the target.
  
The experimental configuration is suitable for one-dimensional modelling, although for more accurate results an explicit geometrical model including source anisotropy, and calculation in time domain is recommended.
  
The recommended (semi) 2-D MCNP5(X) input allowing time and energy domain analysis is provided in NI2d.i. Two older computational models for the MCNP-4B code are also included:
mcnp1d.inp: simple spherical 1-D model with an isotropic source,
mcnp3d.inp: model including sphere geometry, collimators and source anisotropy.

NEA-1553/50
SINBAD-OKTAVIAN/FE
==================
Measurement System and Uncertainties:
------------------------------------
The detectors used were:
  
Detector           Diameter Thickness   Energy covered
                     (cm)      (cm)         (MeV)  
-----------------------------------------------------------------
NE-213 scintillator  12.7      5.08        0.07-14
    
Li-6 glass scint.    12.7      2.54         0.01-1

NE-213 scintillator  3.81      3.81         (14)    neutron source
                                                    monitor
------------------------------------------------------------------
Large statistical errors above about 3 MeV and below about 0.02 MeV, elsewhere 5 - 15%.
Uncertainties in neutron source in 14 to 15 MeV range are about 8%.
  
Description of Results and Analysis:
-----------------------------------
Neutron leakage current spectrum of neutrons was measured in "absolute values" by the time-of-flight technique between 10 keV and 14 MeV, about 9.5 m from the sphere centre. The energy range covers the cross section valley of iron at 24 keV. Because of the presence of the collimators the detectors could not see the entire surface of the sphere, but only the solid angle of 17.28 from the sphere centre.
  
The experimental configuration was modelled originally in one-dimension. The authors found no difference in the results of ANISN calculations using angle dependent shell source and assuming isotropic source. No difference was also found between ANISN P5, S32 and P8, S64 calculations. The measured spectrum was found to be very sensitive to the resonant elastic cross-section (deep penetration neutrons) [8].
For information, two 1-D calculational model for MCNP-4B code are provided.
  
Recently a more realistic (semi) 2D model FE2d.i was developed (Ref. [9]).

NEA-1553/52
SINBAD-OKTAVIAN/SI
==================
Measurement System:
------------------
A cylindrical liquid organic scintillator NE-218 (12.7 cm-diam, 5.1 cm-long) was used as a neutron detector. The detector efficiency was determined by combining:
1) the Monte Carlo calculation,
2) the measured efficiency derived from the TOF measurement of Cf-252 spontaneous fission spectrum and the Watt's spectrum, and
3) the measured efficiency from the leakage spectrum from a graphite sphere, 30 cm in diameter with the similar detection system.
  
To monitor the absolute neutron spectrum per source neutron, a cylindrical niobium foil was set in front of the tritium target and irradiated during the TOF experiment. From the gamma-ray intensity of the induced activity, Nb-92m and the integrated counts of the source neutron spectrum, the absolute neutron leakage spectrum can be obtained.
  
To measure the gamma spectra, OKTAVIAN was run in the pulsed mode with a repetition frequency of 500 kHz. The pulse width was 3 ns in FWHM and the difference in flight times between the 14 MeV neutrons and the prompt gamma-rays was about 90 ns from the sphere to the detector. Since those were enough to separate the gamma-rays from the neutron background in the TOF spectra, the desired gamma-rays could be discriminated from a neutron background.
  
The gamma emission spectra were dominated by the gamma-rays from (n,n') and (n,2n) reactions rather than the gamma-rays from (n,xg) reaction. The data are therefore available to the assessment in the nuclear data for energy distributions of gamma-rays from non-elastic scattering by high energy neutrons.
  
Description of Results and Analysis:
-----------------------------------
The measured neutron leakage spectrum from a 60 cm diameter silicon pile is given. Numerical data for the gamma-ray leakage spectra, measured from a 40 cm anda 60 cm diameter silicon piles are given.
In ref. [9] the neutron leakage spectra measurements from a 40 cm Si file are mentioned.
  
Error Assessment:
  
The experimental errors in the measured neutron spectra include only statistical deviation (1 s). The relative error to measure the niobium activation foils is less than 1 % (0.4 to 1 %), which is not included.
  
In the measured gamma spectra the following sources were included in the errors:
(a) Uncertainty in monitoring absolute fluxes of the source neutrons,
(b) Errors of the response matrix,
(c) Statistical deviation (1 s).
  
Three sets of inputs are provided:
- two older 1-dimensional (1D) MCNP-4B inputs for 60cm Si pile experiment, one model (file mcnp60_n.inp) for neutron, and the second (mcnp60_n.inp) for gamma transport calculations.
- two routine MCNPX(5) (semi) 2-D models in which the neutron source or the gamma source is specified (AL2dns.i, AL2dgs.i).
- Detailed 3-dimensional MCNPX(5) model including the full experimental information for both neutron and gamma spectra (AL3dn.i).
- Detailed 3-dimensional MCNPX(5) model including the full experimental information and explicit D-T source for gamma spectra analysis (AL3dg.i). The D-T source routine (patch_DT) is needed to run this input.
  
Results using the recommended more accurate 2-D and 3-D models are discussed in [10].

Older benchmark calculations with the simple 1D MCNP model and ENDF/B-III silicon data are described in ref. [8]. Measured and calculated spectra are shown in Figures 6, 7 and 8.

Ref. [9] presents the benchmark calculation performed by the MCNP-4A code with JENDL-3.2, JENDL-FF, FENDL-1 and FENDL-2 (ENDF-VI) libraries. Calculated and measured spectra for the 40 cm diameter silicon pile are shown in Figure 9, and for the 60 cm diameter silicon pile in Figures 10, 12 and 13. The results by A. Trkov using the MCNP-4B code and the EFF-3/ENDF/B-VI evaluations are given in Figure 11.

NEA-1553/53
SINBAD-OKTAVIAN-W
=================
Measurement System:
------------------
A cylindrical liquid organic scintillator NE-218 (12.7 cm-diameter, 5.1 cm-long) was used as a neutron detector. The detector efficiency was determined by combining:
1) Monte Carlo calculation,
2) measured efficiency derived from the TOF measurement of Cf-252 spontaneous fission spectrum and the Watt's spectrum, and
3) measured efficiency from the leakage spectrum from a graphite sphere, 30 cm in diameter with the similar detection system.
  
To monitor the absolute neutron spectrum per source neutron, a cylindrical niobium foil was set in front of the tritium target and irradiated during the TOF experiment. From the gamma-ray intensity of the induced activity, Nb-92m and the integrated counts of the source neutron spectrum, the absolute neutron leakage spectrum can be obtained. The formulation of this procedure is described in the Oktavian Report [7].
  
To measure the gamma spectra, OKTAVIAN was run in the pulsed mode with a repetition frequency of 500 kHz. The pulse width was 3 ns in FWHM and the difference in flight times between the 14 MeV neutrons and the prompt gamma-rays was about 90 ns from the sphere to the detector. Since those were enough to separate the gamma-rays from the neutron background in the TOF spectra, the desired gamma-rays could be discriminated from a neutron background.
  
The gamma emission spectra were dominated by the gamma-rays from (n,n') and (n,2n) reactions rather than the gamma-rays from (n,xg) reaction. The data are therefore available to the assessment in the nuclear data for energy distributions of gamma-rays from non-elastic scattering by high energy neutrons.
  
Description of Results and Analysis:
-----------------------------------
The measured neutron leakage spectrum from a 40 cm diameter tungsten pile, numerical data for the gamma-ray leakage spectra, measured from a 40 cm diameter tungsten pile are given. The energy integrated data taken are given in 4 energy groups.
  
Error Assessment:
The experimental errors in the measured neutron spectra include only statistical deviation (1 s). The relative error to measure the niobium activation foils is less than 1 % (0.4 to 1 %), which is not included.
  
In the measured gamma spectra the following sources were included in the errors:
(a) Uncertainty in monitoring absolute fluxes of the source neutrons,
(b) Errors of the response matrix,
(c) Statistical deviation (1 s).
  
Three sets of inputs are provided:
- two older 1-dimensional (1D) MCNP-4B input models, one model (mcnp4b-n.inp) for neutron, and the second (mcnp4b-g.inp) for gamma transport calculations.
- two routine MCNPX(5) (semi) 2-D models in which the neutron source or the gamma source is specified (W2dns.i, W2dgs.i).
- Detailed 3-dimensional MCNPX(5) model including the full experimental information for neutron spectra (W3dn.i).
- Detailed 3-dimensional MCNPX(5) model including the full experimental information and explicit D-T source for gamma spectra analysis (W3dg.i). The D-T source routine (patch_DT) is needed to run this input.
  
Results using the recommended 2-D and 3-D models are discussed in [11].
  
Older benchmark calculations performed with the simple (obsolete) 1D MCNP model and the FSXLIB-J3 and ENDL-85 libraries are described in ref. [8]. Ref. [9] presents the benchmark calculation performed by the 1D MCNP-4A model and the JENDL-3.1, JENDL-3.2, JENDL-FF, FENDL-1, EFF and BMCCS libraries.
Ref. [10] presents the calculations performed at the OECD/NEA DB using the simple 1D MCNP inputs with the FENDL-2 (=JENDL-FF) and ENDF/B-VI.8 evaluations. The calculated and measured spectra for the 40 cm diameter tungsten pile are shown in Fig. 6 and Fig. 7. It is also shown that the fine neutron source spectra given in the MCNP input reproduces better the fusion peak spectra then the one given in Table 2.
The gamma spectra shown in Fig. 8 indicate good agreement between the experiment and the calculation using the ENDF/B-VI.3 data.

NEA-1553/54
SINBAD-FNG-BLKT
===============
Measurement Systems and Uncertainties
-------------------------------------
Activation foils were placed along the center line of the experimental block in
line with the source axis. Foils were placed at various depths along the axis,
irradiated with the neutron source, and its activity recorded from calibrated HPGe
detectors.  Errors from counting statistics, uncertainty of detector calibrations,
and the uncertainty of the source intensity are included with the reaction rate
results.  Other errors are negligible.

A set of TLD-300 detectors were used for measuring the nuclear heating as a
function of depth inside the assembly.  These were calibrated by a 60Co source.

The neutron source intensity was measured by the associated particle method,
using a Silicon Surface Barrier Detector (SSD).  The accuracy of the measurement
was +/- 2%.

Description of Results and Analysis
-----------------------------------
The activation foils were place at 3.43, 10.32, 17.15, 23.95, 30.80, 41.85, 46.85,
53.80, 60.55, 67.40, 74.40, 81.10, 87.75, and 92.15 cm from the front surface of
the experimental block.  Results are given with the total, systematic, and random
error (in %).  MCNP.4A was used to compute reaction rates for the foils.

The absorbed dose measurements from the TLD-300 were taken at 3.36, 10.23, 17.09,
23.95, 30.81, 41.88, 46.88, 53.89, 60.72, 67.58, 74.57, 81.20, 88.01, 95.36, 97.36,
99.76, 101.96 cm from the front surface of the experimental block.  Note that the
measurements included the rear TF coil.  The measurements include systematic,
random and total error (%), with accompanying MCNP calculations.  The absorbed
dose in a material, Q, is given by the following equations:

Q = Qn+Qp
(QTLD) = CpQp+ CeCnQn
Q = QTLD + Qp[(1-CeCn)Qn/Qp+(1-Cp)] ,

where:
QTLD is the response in the TLD,
Qp and Qn is the gamma and neutron absorbed dose in the material,
Cp is the ratio of photon dose in the TLD to the photon dose in the material,
Cn is the ratio of neutron dose in the TLD to the neutron dose in the material.
Finally, Ce is the neutron sensitivity of the TLD-300 taken from the published
in the literature [6].

The cell heating deposition tally (F6 N,P in MCNP) was used to calculate the
neutron and gamma heating in stainless steel or copper during n,p coupled
transport calculations.  The kerma factors for the photons and neutrons were
taken from FENDL.1 and EFF-3 libraries.

NEA-1553/55
SINBAD-FNG-DOSE-RATE
====================
Measurement System:
------------------
The following quantities are measured :

a - Shut down dose rate in the cavity centre (continous measurement by an active dosemeter)
b - Dose rate in the cavity centre, integrated measurement by thermo-luminescent detectors (TLD-300, GR-200A)
c - Ni-58(n,p)Co-58 and Ni-58(n,2n)Ni-57 activation reaction rates during irradiation, using Ni foils

The continuous measurement of the dose rate was taken in the cavity centre after shut down from half an hour to more than three months of cooling time, using a  Geiger-Muller detector (G-M, Mod. 7312 - Vacutec) with a Multi-Channel Scaler with variable dwell time (EG&G Ortec). The detector (12 mm in diameter, 80 mm in length) was located in the cavity centre in front of the open channel (Fig.3). The total experimental uncertainty was ? 10% for G-M detector.

High sensitivity thermoluminescent detectors of the type TLD-300 (CaF2:Tm) [6], GR-200A (LiF:Mn, Cu, P) were also used to measure independently the dose rate in the cavity centre (close to G-M) at four decay times (8.2, 12.4, 19.2 and 33.2 days, for time intervals ranging from 18 to 22.5 hours). The total error associated with the measurements was ?17%. The dose rates measured with TLD in the cavity centre was in agreement within 12% with values obtained with the Geiger-Muller detector, within the combined experimental uncertainties.

Activation measurement were carried out using Ni foils located on the cavity walls (Figure 4). The goal was to measure the reaction rate of Ni-58(n,p) producing the Co-58  (responsible of most of the dose rate in the relevant decay time), and the reaction rate of Ni-58(n,2n)  which produces the Ni-57 (the second most important contributor to total dose rate in the first week after shutdown, after Mn-56 is decayed). The total experimental error was ?5%.

In May 8-10, 2000 the mock-up was irradiated by 14-MeV neutrons at FNG, for a total of 18 hours in three days (Table 5 and Figure 5). The total neutron production was 1.815E+15.


Description of Results and Analysis:
-----------------------------------
The experiment analysis was performed using a rigorous two-step method (R2S) employing the MCNP-4C [7] code with FENDL/MC-2.0 [8] cross sections for calculating neutron transport (in a first run) and decay gamma transport (in a second run) in sequential order, and the FISPACT [9] inventory code with FENDL/A-2.0 [10] activation cross sections for calculating the decay gamma source distribution as a function of irradiation history and cooling time.

Two different MCNP models of the FNG assembly were employed: one for the neutron transport calculation during irradiation (mcnp_n.inp, ?irradiation model?) and the other one for the decay gamma transport calculation after irradiation (mcnp_g.inp, ?shut-down model?). In this way  proper account is taken of the fact that during the irradiation the central cavity was empty and the lateral access was plugged whereas after irradiation the plug was removed and the detectors were inserted into the cavity.

The neutron flux spectra are calculated in the VITAMIN-J 175 group structure for all non-void cells of the ?FNG irradiation model? and are routed to FISPACT. Activation inventories and decay gamma sources (spectrum and intensity) are then calculated for all material cells making use of the associated neutron flux spectra. This requires one FISPACT-calculation per cell and material taking into account the proper irradiation history.

The resulting decay gamma source distribution is then routed back to MCNP. The MCNP decay gamma transport calculation is performed with the ?FNG shut-down model? (mcnp_g.inp) making proper use of the decay gamma sources as provided by the preceding FISPACT calculations for all non-void geometry cells. The dose rate in air is calculated in a cell in the cavity centre (cell#651 simulating the GM detector) using tally f6 of MCNP.

The description of the irradiation history is given in the FISPACT input fisp_620.inp (relative to one cell, e.g. cell#620 of mcnp_n.inp).

A pre-analysis was carried out in order to investigate the origin of the doserate: Figure 6 shows the contributions of major nuclides to the total contact dose rate, as calculated by FISPACT at the inner cavity wall. Mn-56 dominates at short times (i.e. t<1 d), Ni-57 at around 1 d, and then Co-58 dominates in the time range of practical interest for allowing personal access for maintenance purposes. The nuclei  considered in the figure contribute to more than 95% of the total dose rate, as shown in the same figure by the black line.

The Ni-58(n,p)Co-58 or Ni-58(n,2n)Ni-57 reaction rates were calculated in two ways:

1. using a procedure similar to R2S method, i.e. using FISPACT with Ni-58(n,p) and (n,2n) cross sections from FENDL/A-2. Statistical errors on MCNP flux calculations are ?2.5%.
2. calculating the reaction rate is directly in the MCNP run taking the Ni-58(n,p) and (n,2n) cross sections from the dosimetry file IRDF-90.2 [11] and from FENDL/MC-2. Statistical errors on  reaction rate calculations are ?2.5%.

The measured dose rate are given in Table 6 (G-M) and 7 (TLD), and Figure 7. The calculated ones are given in Table 10 at cooling times equal to 1, 7, 15, 30, 60 days.
The measured Ni-58(n,p)Co-58 and Ni-58(n,2n) reaction rates are given in Tables 8 and 9 respectively. The calculated reaction rates are given in Tables 11-12.

NEA-1553/56
SINBAD-FNG-SIC
==============
Measurement System:
------------------
The following quantities are measured :

    a - Neutron reaction rates by activation foils
    b - Nuclear heating by thermo-luminescent detectors (GR-200A)

Four different reactions: 197Au(n,g), 58Ni(n,p), 27Al(n,a) and 93Nb(n,2n) were used to derive the neutron flux, from thermal energy up to the fusion neutron peak. The reaction rates were measured at four experimental positions, 10.41 cm, 25.65 cm, 40.89 cm and 56.13 cm respectively from the block surface, using the radiometric techniques based upon the use of absolutely calibrated HPGe detectors. The overall contribution to the quoted uncertainty comes from the HPGe calibration (?2%), measured activity (
Nuclear heating was measured at 14.99 cm, 30.23 cm, 45.47 cm and 60.71 cm depth from the block surface using high sensitivity GR-200 (LiF:Mg,Cu,P) thermoluminescent dosimeters (TLD). Since the TLD technique is a relative method, TLDs were calibrated against a Co-60 g-ray source. The calibration uncertainty was within ?5%. The calibration was performed using a secondary standard. Four TLDs were located in each experimental position and the averaged values of the measured TL light was used as experimental data. From this datum, after conversion via the calibration factor, the experimental total dose in TLDs was obtained. The total uncertainty is ?7%. This uncertainty includes the above mentioned calibration error, the error (standard deviation) on the averaged TL light value, the uncertainty on the FNG neutron yield (?3%), summed using the quadratic law. The experimental results are given in Table 4.


Description of Results and Analysis:
-----------------------------------
The experimental results (E) were analysed by the Monte Carlo code MCNP-4C [6] using point-wise cross sections derived from FENDL-2.0 [7], EFF-2.4 and also with the new evaluation for 28Si included in EFF-3.0. The calculations are presented in [1] and [2]. Activation reaction rates were calculated using tally f4 of MCNP, the dosimetric reactions needed for the calculation were taken from IRDF-90.2 library [8] (mcnp.inp). The calculated reaction rates (C) are given in Table 5 together with the statistical uncertainty.

The gamma dose Dg  and the neutron dose Dn in TLDs was calculated with tally f6 of MCNP, using a very detailed model in MCNP of the dosimeters and of their plastic containers (mcnp_tld.inp). The dose in TLD was obtained from:

DTLD = Kn*Dn + Dg

where Kn is neutron sensitivity of the TL detector to neutrons and is usually depending upon the neutron energy. The Kn data were taken from [9] and were then weighted over the calculated neutron spectra at each experimental position. The resulting  Kn values are reported in Table 7. Comparison between the measured dose in TLD and the same quantity calculated with EFF-2.4, EFF-3.0 and FENDL-2.0 nuclear data is also given in Tab. 7.

The transport and cross section sensitivity/uncertainty analyses were also performed using the deterministic codes DORT, TWODANT and SUSD3D, and are presented in [3] and [4]. Analytic uncollided and first collision source approach was used in order to mitigate the ray effects.
Input data for the following codes are included:
- TRANSX: nuclear cross section preparation;
- GRTUNCL and DORT: uncollided/first collision source and discrete ordinates (SN) transport calculation;
- TWODANT: SN neutron transport using first collision approach.

NEA-1553/57
SINBAD-FNG-SS
=============
Measurement Systems and Uncertainties:
-------------------------------------
The accuracy of the source measurement lay between ?4.4% and ?1.6% using a SSB detector which detects associated particles from the T(d,n)He reaction.  Errors are reported for gamma-ray interference and uncertainty in the three HPGe detectors.  The systematic error is the uncertainty on the neutron source intensity. Calibration of the HPGe detectors were within ?2% uncertainty. Activation reactions and nuclear data employed for detector foils follows:

Reaction          Half-life   Isotopic      g-ray         g-ray
                              abundance     energy      branching
                                (%)         (keV)          (%)
27Al(n,a)24Na       14.96h      100.0       1368.6       100.0
56Fe(n,p)56Mn       2.577h      91.72       846.8         98.87
58Ni(n,2n)57Ni      1.503d      68.27       1377.6        80.0
58Ni(n,p)58Co       70.92d      68.27       810.8         99.44
115In(n,n')115mIn   4.486h      95.7        336.24        45.9
55Mn(n,g)56Mn       2.577h      100.0       846.8         98.87
197Au(n,g)198Au     2.696d      100.0       411.8         95.56

Description of Results and Analysis:
-----------------------------------
The foils were placed at various depths along the central z-axis, from 5 cm to 60 cm.  The background was determined to be between 4% and 8%.  IRDF-90 Covariance data in 175 energy group format was used to determine the error on reaction rates due to uncertainty of the activation cross sections.  Transport cross sections were used from EFF.1 and the activation cross sections for the foil detectors were taken from IRDF.90 with the exception of 55Mn(n,) which was taken from EFF.2.  Data is presented by foil depth in cm, measured reaction rate, 1024 reactions/(source neutron), random error, systematic error, and the total error (%).

NEA-1553/58
SINBAD-FNG-STREAMING
====================
Measurement System:
------------------
The following activation reactions were selected to measure the neutron flux:

Reaction   Effective Threshold
                    (MeV)
93Nb(n,2n)92Nb   10.8
27Al(n,a)24Na     8.5
58Ni(n,p)58Co     2.9
197Au(n,g)198Au     -

All foils had a diameter of 18 mm. 1-3 mm thick foils were used for Nb, Al and Ni detectors (1 mm up to the depth of 38.65 cm and in the cavity, 2 mm behind the cavity between 46.35 and 73.90 cm, and 3 mm in the last 3 positions). For 197Au(n,g)198Au reaction 0.05 mm thick foils were used.

Inside the channel the foils were located in axis with the channel, at y = 0.25, 12.95, 25.95, 38.65 cm  (foil centers) from the shielding block surface. Eleven activation foils were located inside the cavity to monitor the neutron flux gradient and the effect of the void channel. The foils were located in the positions shown in Fig. 4, between y = 39.12 cm and y = 43.82 mm depth (foils centers). Eight activation foils are located behind the cavity, at depths:
y = 46.35, 53.30, 60.05, 66.90, 73.90, 80.60, 87.25, 91.65 cm (foil centers).

Nuclear heating was measured in the shielding assembly behind the channel/cavity, in the same positions as for activation foils, and inside the superconducting magnet, using CaF2:Tm thermo-luminescent detectors (TLD-300) of size 0.32 x 0.32 x 0.09 cm3 provided by Harshaw Company. The measured absorbed dose in TLD-300, Q-TLD (E) is obtained from the peak 3 response in the glow curve relative to Co-60 calibration. The air kerma was converted into absorbed dose in TLD-300 using the photon energy attenuation coefficient from [2].


Description of Results and Analysis:
-----------------------------------
Measured neutron reaction rates are given in Tables 5-7. They  were compared with the same quantities calculated with the  same tools used in the ITER design, i.e. MCNP 4A/B [10] code and FENDL-1/FENDL-2 [11,12] nuclear data libraries. Analysis was performed also using the EFF-3 [13] library. Dosimetric cross sections were taken from IRDF-90 [14] to compute the neutron reaction rates in the MCNP run. The geometry model for MCNP-4B used in the calculation of reaction rates, including neutron source backing and the experimental environment (walls, floor, racks, ...) is given in mcnpfoil.inp. C/E values were provided for all measured reaction rates, they are reported in Tables 9-11.

Measured absorbed dose in TLD-300 is given in Table 8. It is compared with the same quantity calculated with  MCNP code and FENDL-1/FENDL-2, EFF-3.1 nuclear data libraries.
The geometrical model for MCNP used in the calculation of nuclear heating, is given in mcnp_nh.inp. The comparison with TLD measurements requires, however, the calculation of the absorbed dose in TLD, taking into account the effects of the electron transport at the interface between the TLD and the surrounding material. The absorbed dose in TLD (QTLD) is related to total dose in material Q according to the following expressions:

                      Q = Qn+Qp
                      (QTLD ) =  Cp Qp + Ce Cn Qn     (1)

where QTLD is the absorbed dose in TLD,
Qn, Qp is the absorbed dose due to neutrons, photons in material (i.e. steel or copper),
Cn is the ratio of the TLD/material absorbed neutron dose,
Cp is the ratio of the TLD/material absorbed photon dose,
Ce is the TLD neutron dose efficiency with respect to the photon dose ciency, taken from published data and weighted over the neutron spectra.

Simplified models have been adopted representing a smaller region limited to the TLD volume and to a sufficient amount of surrounding material, to calculate Qn, Qp, Cn and Cp. A neutron or photon surface source is applied to the boundary of the limited region in the simplified models, with energy spectra recorded from the mcnp_nh.inp run. These models are given in mcnp_hss.inp, mcnp_hcu.inp and mcnp_tld.inp for steel, copper and TLD material respectively.

The Cn, Cp, Ce factors and Qn, Qp calculated using FENDL-1, FENDL-2 and EFF-3.1 are used to derive the calculated dose in TLD according to Eq.1. For all calculated absorbed doses the fractional standard deviation is about 5% or less. C/E values
are given in Tables 12.

The calculations performed using the 2D discrete ordinates transport method is described in [5] and [15]. The input data are also provided, including inputs for the codes TRANSX, GIP, GRTUNCL and DORT.

NEA-1553/59
SINBAD-FNG-W
============
Measurement System:
------------------
The following quantities are measured :

a - Neutron reaction rates by activation foils
b - Nuclear heating by thermo-luminescent detectors (TLD-300)

Eight different reactions: 197Au(n,g), 55Mn(n,g), 115In(n,n?), 58Ni(n,p), 56Fe(n,p),  27Al(n,a), 58Ni(n,2n), 90Zr(n,2n) and 93Nb(n,2n) were used to derive the neutron flux, from low energy up to the fusion neutron peak. The reaction rates were measured in four experimental positions at different depths from the block surface, using the radiometric techniques based upon the use of absolutely calibrated HPGe detectors. During the activation foil measurements, the lateral access channels were completely closed by means of 4 ad hoc cylinders made of DENSIMET ? 180, a thin slot was realised (4.4 mm) to locate activation foils in position, using a thin Al holder. The foils were irradiated in three irradiations: in the first one Zr, Al and Mn foils were irradiated, the foils arrangement is described in the MCNP input file ZrAlMn.mcp; in the second one Nb, Ni and Au foils were irradiated, the foils arrangement is described in NbNiAu.mcp; in the last irradiation Fe and In foils were irradiated, the foils arrangement is described in FeIn.mcp. The experimental results are given in Table 4.

Gamma heating was measured using TLD-300 dosimeters (CaF2:Tm). TLDs calibration was performed using Co-60 secondary standard, from 50 mGy up to 4 Gy in air, converted into absorbed dose in TLD-300 using the photon energy attenuation coefficients from Hubble [4]. Seven TLDs chips (3.2x3.2x0.9 mm3 each) were located in each experimental position, using the same experimental arrangement as for the activation foils, and enclosed in a perspex  holder 1 mm thick. The experimental arrangement is described in the MCNP input file mcnp_tld.inp.The measured dose in TLD-300 is given in Table 5.

Description of Results and Analysis:
-----------------------------------
The experiment was analysed by using the Monte Carlo code MCNP-4C [2, 5] using for W, Fe and Ni the point-wise cross sections derived from EFF-2.4 and FENDL-2.0 [6]. In the case of EFF calculation, the Fe cross section were taken from EFF-3.0. The MCNP model of the experimental set-up was is given in ZrAlMn.mcp, NbNiAu.mcp and FeIn.mcp for the activation foils measurements, and in mcnp_tld.inp for the TLD measurements. The track length estimator was used (tally 4 of MCNP) for fluxes and reaction rates calculation, while the gamma heating is calculated from the gamma energy deposition over the TLD cells (tally 6 of MCNP). All dosimetric reactions needed for the calculation of reaction rates were taken from IRDF-90.2 library [7]. The calculated reaction rates and gamma doses (C) are given in Tables 6 and 7 respectively, together with the MCNP statistical uncertainty.

Deterministic transport and cross section sensitivity/uncertainty analyses using the DORT, TWODANT and SUSD3D codes are presented in [8], [9] and [10]. The following input data used in these analyses are included here:
- TRANSX (cross section preparation),
- GRTUNCL and DORT (uncollided/first collision source and transport calculation)
- TWODANT (neutron transport using first collision approach).
The 2-dimensional cylindrical geometry model used in the DORT and TWODANT deterministic transport calculations is shown in Figure 10.

NEA-1553/60
SINBAD-FNS-DUCT
===============
Measurement System and Uncertainties:
-------------------------------------
Neutron spectra above 2 MeV were measured at several positions by a spherical NE213 scintillation spectrometer 40 mm in diameter. The two-gain method was adopted in the electronics system to make a wider energy range available. Neutron and gamma-ray signals were separated by a pulse shape discrimination technique based on the differences in rise time of the signal. The pulse height spectrum of recoil protons which represent neutron events was unfolded with the FORIST code [2] to obtain the neutron energy spectrum. The reactions of 93Nb(n,2n)92mNb, 115In(n,n')115mIn and 197Au(n,g)198Au were employed as neutron activation dosimeters. The first reaction is effective to evaluate the 14 MeV neutron flux, the second is sensitive to fast neutrons above 1 MeV, and the third helps to understand the amount of the thermal and epithermal neutron flux.

Description of Results and Analysis:
------------------------------------
Reaction rates measured with activation foils in the bent duct and on the back surface of the assembly are shown in Table 1 and Fig. 4 under the same normalisation as the spectrum. It is observed that the reaction rates of 93Nb(n,2n)92mNb and 115In(n,n')115mIn caused mainly by fast neutrons prominently decrease after the duct bends, while those of 197Au(n,g)198Au do not show clear change around the bends.

Measured neutron spectra are shown in Table 2 and Fig. 3. The D-T neutron source intensity was normalised to unity. Positions #3, #5 and #7 are located in the duct (see Fig. 2). The spectra become softer with increasing path along the duct from the inlet. The spectrum at position #9 is higher than at position #7, because the position #9 is located on the extension of the first leg and the shield between positions #3 and #9 is only 50 cm. It is noteworthy that the 14 MeV peak at position #9 is even larger than that at position #7.

The experiment was analysed by the Monte Carlo codes MCNP-4B and -4C [3] using the nuclear data libraries FENDL/2 [4] and JENDL-3.3 [5]. The agreement between the calculated and measured values is generally within the Monte Carlo statistical errors, proving that the codes and the nuclear data libraries are sufficiently reliable and accurate to estimate streaming effects in the shielding design of fusion reactors.

NEA-1553/61
SINBAD-FNS-OXYGEN
=================
Measurement System:
------------------
The neutron spectra were measured with the time-of-flight (TOF) technique. A cylindrical NE213 scintillator (50.8 mm-diameter, 50.8 mm-long) was used as a neutron detector. The aperture size of the collimator was such that the scintillator detected only neutrons leaking from the central area of the slab surface behind the source. The scintillators were located about 7 m (see Table 1) from the cylindrical slab and at angles of 0, 12.2, 24.9, 41.8, and 66.8 degrees with respect to the deuteron beam axis. The energy range between 0.05 and 15 MeV was measured by combination of two bias schemes described in [2]. The details of the experimental set-up are shown on Figure 2.


Description of Results and Analysis:
-----------------------------------
The measured angular neutron flux spectra are given in Table 3. The spectra were calculated from the following expression (see Figure 3 for definitions):

F(omega,E)=C(E)/(eps(E).dOmega.As.Sn.T(E))

where C(E) is detected neutron counts, eps(E) the neutron detection efficiency, dOmega the solid angle subtended by the detector to the point on the center of slab surface (=Ad/L*L, where Ad=detector area and L=distance from the slab surface to the detector), As is the effective measured area defined by the detector-collimator system on the plane perpendicular to its axis, Sn number of source neutrons emitted from the D-T target, and T(E) the attenuation correction due to scattering in air.

Error Assessment:
The experimental errors in the measured neutron spectra include only statistical deviation (1 sigma).

Example of Experiment Analysis:
Several sets of inputs are provided:
- Recommended MCNPX(5) input for time domain analysis is given in TOF.i.
- An older input for the MCNP-4B code from ref. [10] is given in file mcnp4b.inp. The calculations were performed using ENDF/B-VI Rev.1 data from the original MCNP library. The results for the neutron spectrum calculation are compared with the measurements in Figures 5 and 6 for different angles.
- The sample input for the DOT-3.5 code taken from ref. [9] is given in file dot35.inp. For the description of the calculational procedure see refs. [4] and [8].

NEA-1553/62
SINBAD-FNS-SKYSHINE
===================
Measurement System:
------------------
Neutron dose rates were measured with the spherical rem-counters, He-3 and BF3 Fuji Electric NSN10002, just on the ground at all measurement points shown in Figure 2.

The neutron spectra were evaluated with boner spheres with multi-moderators of 0.5, 1, 2, 3, 5 and 9 cm thickness along the north direction, and unfolded by SAND-II code.

Secondary gamma-ray spectra were measured with a high purity Ge detector and a NaI scintillation counter with a size of 125 cm in diameter by 125 cm long at distances of 20, 50 and 100 m, and with a NaI scintillation counter with a size of 200 cm in diameter by 200 cm long at distances of 200 and 300 m.


Description of Results and Analysis:
-----------------------------------
The measured neutron dose in the south and south west directions are shown in Figure 3. The numerical data of the neutron dose are given in Table 1. Neutron spectra measured at a distance of 50, 100 and 200 m and unfolded by SAND-II code are shown in Figure 4. They are compared with the first MCNP-4B calculations (forest was not included in the calculational model).

Figure 5 shows the pulse height spectra measured with the NaI scintillation counter at the distances of 20, 50 and 100 m. Numerical data of Fig. 5 are given in the file sky-NaI.txt. The information on the detector response function and the detector efficiency, needed for the comparison with the calculation, is not yet available.

Few characteristic gamma-ray lines can be observed in the spectra, like the natural gamma-ray background from K-40 and Tl-208, the energy peaks from the H-1(n,gamma) and Fe-56(n,gamma) reactions. A continuum spectrum in the energy range of 3-7 MeV was enhanced by the shyshine, which is considered to be the convolution of Compton spectra from some neutron capture reactions. The high purity Ge detector observed additional peaks of the Si-28(n,gamma) reaction at 3.539 and 4.934 MeV at the distance of 20 m only, which is considered to arise from the FNS building.

Experiment Analysis:
Transport calculations with the MCNP-4B code are presented in [1]. The calculations were performed using the JENDL-3.2 nuclear data library. The corresponding sample input for the MCNP-4B code is given in the file mcnp4b-sky.inp. The forest was not modeled in this calculation.

More recent analysis with the MCNP-4C code and JENDL-3.2 nuclear data are described in [2]. The corresponding MCNP-4C input data are included in mcnp4c-sky.inp and the comparison of the measured and calculated neutron dose rate is presented in sky-mcnp4c.xls. In this model the FNS building and the measurement field were modelled in a simplified cylindrical geometry, and the pine forest was modelled by homogenised cell with a height of 10 m.

NEA-1553/69
SINBAD-TUD-FNG-BS
=================
Measurement System:
------------------
A NE213 scintillator was employed for simultaneously measuring the neutron spectra for energies E>1 MeV and the photon spectra for energies E>0.2 MeV. For each registered event both the pulse-height and a pulse-shape parameter were recorded to distinguish between neutrons and photons. Pulse-height distributions from three different hydrogen-filled proportional detectors, one methane-filled proportional detector and a stilbene scintillator were used for determining the neutron flux spectra for energies ranging from 20 keV up to about 3.6 MeV, overlapping with the NE213 spectra [2-6].
Measurement uncertainties are provided with the tabulated spectra and are between 2 and 10 %.
  
Description of Results and Analysis:
-----------------------------------
Neutron energy spectrum:
The NE213 pulse-height spectra were unfolded by the DIFBAS code [7] with a response matrix developed at Physikalisch-Technische Bundesanstalt Braunschweig [8], to obtain spectral fluxes. The evaluation procedure of the proton recoil spectra from the proportional detectors and from the stilbene scintillator consisted in an iterative differentiation with inclusion of the following corrections: wall effect, non-linear light output function, anisotropy effect of stilbene, energy dependent sensitivity of the stilbene crystal and correction of neutron reactions on carbon resulting in alpha-particles. The neutron energy spectra represent a combination of the partial spectra obtained with the different detectors which were in each case corrected for material and size of the detector to a spherical detection volume of 2.0 cm radius filled with SS316.
  
Photon energy spectrum:
The pulse-height distribution from the NE213 detector for photons was unfolded with the DIFBAS code [7] with a response matrix calculated with the MCNP code. Also the photon energy spectra represent the spectral fluences in a SS316 sphere with radius of 2 cm and were normalized to one source neutron.
  
Calculations:
Examples of calculations carried out with the 3-D Monte Carlo code MCNP-4A [9] and the data libraries FENDL-1 [10] and FENDL-2 [11] are presented [12].
The geometry model for MCNP-4A including neutron source backing and the experimental environment (walls, floor, racks, ...) is given by the input file as well as a FORTRAN subroutine for MCNP source description.
  
The input data for the 2D discrete ordinates transport calculation [13] are provided, including inputs for the codes TRANSX, GIP, GRTUNCL and DORT.

NEA-1553/70
SINBAD-TUD-FNG-SIC
==============
Measurement system:
-------------------
Neutron and photon pulse-height spectra were measured simultaneously using an NE 213 scintillation spectrometer. The dimensions of the cylindrical active volume of the detector were 3.8 cm in both height and diameter. Its material had a mass density of 0.874 g/cm3 and an elemental composition of 54.8 at-% H and 45.2 at-% C. The scintillator was coupled to a photomultiplier by means of a 50 cm long light guide. When the detector was located at one of the positions, the other ones were filled with pieces of SiC.
  
Description of results and analysis:
------------------------------------
Data evaluation:
The DIFBAS code developed at PTB Braunschweig [2,3] was employed for unfolding the measured pulse-height distributions in order to generate the neutron and photon flux spectra. They were obtained as absolute fluxes, as the response matrices have been determined on an absolute scale by detailed simulations of experimental distributions from mono-energetic neutron and photon sources with Monte Carlo codes.
  
Calculations:
A computational analysis was performed with the Monte Carlo code MCNP-4C [4] using a full 3D model of the assembly, the neutron generator and the experimental hall. Nuclear data were taken from the FENDL/MC-2.0 data library [5] except for Si-28, for which EFF-3.0 [6] data were used. The spectra were calculated as average flux in the scintillator volume by means of the track length estimator of MCNP.

NEA-1553/71
SINBAD-FNG-HCPB
===============
Measurement System:
------------------
The following quantities are measured:
  
a - Neutron reaction rates by activation foils placed in the beryllium central layer
b - tritium production by Li2CO3 pellets (containing natural lithium) in the double ceramic layers
c - Nuclear heating by thermo-luminescent detectors (TLD-300) in the double ceramic layers  
  
Four reactions, 197Au(n,g), 58Ni(n,p), 27Al(n,a) and 93Nb(n,2n) were used to derive the neutron flux, from low energy up to the fusion neutron peak. The reaction rates were measured at four experimental positions at about y=4.2, 10.5, 16.8 and 23.1 cm from the block surface, using the radiometric techniques based upon the use of absolutely calibrated HPGe detectors. During the activation foil measurements, the lateral access channels were completely closed by means of 4 ad hoc cylinders. The arrangement of the foils for the activation measurements is described in the MCNP input file mcnp-hcpb.i and shown in Figure 7.  
  
The tritium production was measured using the 6Li(n,t) and 7Li(n,t) reactions, covering respectively fast ant thermal neutron energies. The detectors were placed in the removable tubes (see hcpb-5.jpg) at four positions along the central beam axes of the block at about y=4.2, 10.5, 16.8 and 23.1 cm from the front surface of the mock-up. Altogether 16 measurement positions were available.
The experimental results are given in Table 4.  
  
Description of Results and Analysis:
-----------------------------------
The experiment was analysed by the Monte Carlo code MCNP-4C and the deterministic codes DORT and TORT using the cross sections derived from EFF-3.1 and FENDL-2.1. All dosimetric reactions needed for the calculation of reaction rates were taken from the IRDF-90.2 and -2002 libraries [7].
  
The MCNP model of the experimental set-up is given in mcnp-hcpb.i. The track length estimator was used (tally 4 of MCNP) for fluxes and reaction rates calculation.
  
The deterministic transport and cross section sensitivity/uncertainty analyses using the DORT, TORT and SUSD3D codes are presented in [10], [11] and [12]. The following input data used in these analyses are included here:
- TRANSX (cross section preparation),
- GIP (cross section preparation),
- GRTUNCL and DORT (uncollided/first collision source and 2D transport calculation)
- GRTUNCL3D and TORT (neutron 3D transport using first collision approach).
2-dimensional (2D) cylindrical and 3D geometry models were used in the DORT and TORT deterministic transport calculations, respectively.

NEA-1553/72
SINBAD-FNS-C-CYLIND
===================
Measurement System:
------------------
Fission rate were measured by micro-fission chambers of 235U, 238U, 232Th and 237Np, and fission track detectors.
  
Reaction rate distribution along the central axis of the assembly were obtained by foil activation technique. The Al, Ni, Zr and Nb foils were 10 mm in diameter and 0.5 mm in thickness; the In foils were 10 mm square and 0.1 mm thick; the Au foils were 10 mm square and 0.001 mm thick.
  
The neutron spectra were measured by a small sphere (14 mm diam.) NE213 organic scintillation counter at 8 positions. The proton recoil spectrum was unfolded by the FORIST code.
Gamma-ray dose rates were measured by thermoluminescence dosimeters (TLDs).
  
Description of Results and Analysis:
-----------------------------------
Fission rates were measured by the micro-fission chambers. Reaction rate distribution obtained by the foil activation technique is shown. Decay data needed for reaction rate calculation and error analysis for the reaction rate measurement are summarized. Response rates of TLDs are shown. Fission rates measured by the fission track detector method are shown. Fission rates distribution in the graphite assembly measured by fission track detector method and those by micro-fission chambers are shown. Neutron spectra measured by the NE213 spectrometer at eight positions are given. Detector loction and alpha count of the spectrum measurements are shown.
  
Error Assessment:
Some descriptions about error assessment are given in the Tables.
Major sources of error of fission chamber measurements were uncertainties in the neutron yield (2.1 %), effective fissile atom numbers (3.2 ~ 4.6 %) and positioning (1 %).
The sources of error for the reaction rate measurements are listed.
The errors of the unfolded spectra include two contributions. One of them, the statistical error, is included in the tables together with the spectra.
The other error is related to the response matrices which are used in the unfolding procedure. This is included in the systematic errors summarized.

NEA-1553/73
SINBAD-FNS-V
============
Measurement System:
------------------
Two experimental channels for insertion of detectors were set on the lateral surface of the assembly, and detectors were placed on the central axis of the assembly at three positions: one on the front surface and other two inside the assembly at the depths of 76.2 and 177.8 mm. The report [1] states that the spectra measurements were performed one by one, we conclude therefore that the other experimental channels were closed (plugged with vanadium). The dosimetric foils were irradiated together and simultaneously in plural detector channels.

No information on the background correction was found in the literature.

Six techniques were employed to measure neutrons and gamma-rays. The detailed description of the experimental techniques is given in [1] (pages 5-20):

(1) 14 mm-diameter spherical NE213 liquid organic scintillator was used as the fast (above 2 MeV) neutron spectrometer. It was inserted into the experimental channel hole of 22 mm diameter.
Various sources of the NE213 measurement uncertainties are discussed in [1] (page 5) and the systematic errors are summarized in Table 2.

(2) A pair of proton recoil gas proportional counters (PRC) was used to measure neutron spectrum in the 20 keV to 1 MeV energy range. The counter has a cylindrical shape with the outer diameter of 19 mm and the effective length of 127 mm. The counter is inserted into an experimental hole of 21 mm in diameter with its pre-amplifier. PRC is made of type 304 stainless steel of a thickness of 0.41 mm.

Two types of counters were used to measure neutron spectrum in a wide energy range. Hydrogen gas at 0.5677 MPa (5.789 kgf/cm2) with 1 percent of CH4 filled counter was used for low-energy neutrons from 3 keV to 150 keV while for the high-energy neutrons from 150 keV to 1 MeV the counter was filled with 50-50 mixture of hydrogen and argon gases with 1.8 percent of nitrogen at 0.6102 MPa (6.222 kgf/cm2).

Error Assessment: Possible error sources are gas pressure (number of hydrogen atom), n-p scattering cross section, fitting error for differentiation of recoil proton spectrum due to count statistics and calibration of recoil proton energy. The fitting error is the largest, ~ 3-10 % above 10 keV, while the other errors are expected to be less than 1%. Neutron spectra below 10 keV tend to become smaller due to the uncertainty of the W-value, which is the average energy loss per ion pair. The error due to W-value is not included in the experimental errors.

(3) Slowing down time (SDT) method for the neutron spectrum from 1 eV to 300 eV. A BF3 gas proportional counter with an outer diameter of 14 mm and an effective length of 99 mm, containing 96 % boron-10 (B-10) enriched BF3 gas at the pressure of 71.5 kPa, was used for neutron detection. Using the standard thermal neutron field the effective number of B-10 atoms in the counter was determined to be 2.18E20 +- 3 %. The counter was inserted into one of the experimental holes of the experimental assembly. Experimental uncertainties associated with the measured spectra are summarized in Table 3. The overall experimental uncertainties are dominated mostly by the uncertainties of the energy calibration curves. The lowest and highest energies calibrated are 1.4 eV and 579 eV. In the energy ranges outside the calibration energies, roughly below 1 eV and above 1 keV, the adjusted calibration curves are used.

(4) Dosimetry reaction rate of the Al-27(n,alpha)Na-24, Nb-93(n,2n)Nb-92m, In-115(n,n?)In-115m and Au-197(n,gamma)Au-198 reactions were measured by the foil activation method. Table 4 provides the characteristics for the reactions used.
Typical Al and Nb sample size was 10 mm in diameter and 1 mm in thickness. Indium foils had dimensions of 10 x 10 x 1 mm3. In order to minimize the self-shielding effect for the Au-197(n,gamma) reaction, gold foils with a size of 10 x 10 x 0.001 mm3 were adopted.

Experimental Error and Uncertainty: Major sources of the error for the reaction rate were the gamma-ray counting statistics (0.1 ~ several %) and the detector efficiency (2 ~ 3 %). The error for sum-peak correction was estimated less than 2 % depending on the decay mode and fraction of multiple gamma-ray cascade. The error for the decay correction was reflected from the error of half-life of the activity. If the half-life was accurate, the error for the saturation factor should be less than 1 % even for the short half-life activities.
The other errors associated with foil weight, gamma-ray self-absorption, irradiation time, cooling time and counting time were negligibly small. The error for neutron yield was estimated to be 2 %. The overall error for the major part of reaction rate ranged between 3 ~ 6 %.  Some data for high threshold reaction in the deep positions suffered from poor counting statistics due to low activation rate.

(5) Prompt gamma-ray spectrum were measured by a 40 mm diameter spherical BC537 liquid organic scintillation counter. Outer diameter of the detector is 48 mm and length including a photomultiplier assembly is 262 mm. Experimental uncertainty: As explained in (6) below, typical experimental uncertainties of the gamma-ray heating rate measured by TLDs range between 7 ~ 15 %. The gamma-ray spectra are normalized to the gamma-ray heating rates. Uncertainties of about 10 % is introduced by the normalization procedure. All of the rest of experimental uncertainties, such as statistical errors, uncertainties of the response functions, determination of source D-T neutrons and subtraction of decay gamma-rays, are less than 5 %.  Therefore, total experimental uncertainties are approximately 15 ~ 20 %. Note that only the statistical erorrs are included in the Table 10.

(6) Gamma-ray heating rate were measured by thermoluminescent dosimeters(TLDs) in combination with the atomic number interpolation method. Gamma-ray heating rates of vanadium were deduced by interpolating the gamma-ray heating rates measured by three types of TLDs, Mg2SiO4 (MSO, effective atomic number Zeff = 11.1), Sr2SiO4 (SSO, Zeff = 32.5) and Ba2SiO4 (BSO, Zeff = 49.9).

Sources of error in the measured gamma-ray heating rates are as follows:

Statistical deviation of four TLDs      5 - 15 %
Number of neutrons generated            2 - 3  %
Calibration of the TLD reader             5    %

Due to the subtraction of target gamma-rays and neutron response, the following uncertainties are to be added to the above errors according to the quadratic propagation of error.

Response functions for neutrons    30 %
Neutron energy spectra             10 %
Target gamma-ray                   20 %

Overall errors for the obtained gamma-ray heating rates of vanadium are ~ 10%, except at the positions in the front surface of the assemblies where they are 20 ~ 25 %.


5. Description of Results and Analysis:
---------------------------------------
The measured neutron spectra by the three methods at the depths of 76 mm and 178 mm are shown in Figs. 3 and  4, respectively. The numerical data of the spectra measured by the NE213, PRC and SDT methods are given in Tables 5 - 7, respectively. Table 8 summarises integral neutron fluxes derived from the measured neutron spectra.
The measured dosimetry reaction rates are shown in Table 9.
The measured gamma-ray spectra at the positions of 76 mm and 178 mm are shown in Figs. 5 and  6, and the numerical data of the spectra are shown in Table 10.
Table 11 summarises the measured gamma-ray heating rates in vanadium.

Example of Experiment Analysis:
Several sets of inputs are provided:
- Recommended MCNPX(5) model including the DT source routine (REALISTIC.i).
  Input for MCNP mesh-based variance reduction is given in wwinp_REALISTIC.
  The D-T source routine (DT_MCNP5.TXT) is needed to run this input.
- An older input for the MCNP-4A code is given in file mcnp-v.inp. The
  input was taken from [1], except that the 0 degree source was used instead
  of the source subroutine (not provided in the document [1]).
  Transport calculations with the MCNP-4A code and the JENDL-FF, JENDL-3.2,
  ENDF/B-VI and EFF-3 nuclear data libraries are presented in [2].

NEA-1553/74
SINBAD-FNS-W
============
Measurement System:
------------------
Three experimental channels for insertion of detectors were set on the lateral surface of the assembly, and detectors were placed on the central axis of the assembly at four positions: one on the front surface and other three inside the assembly at the depths of 76, 228 and 380 mm. The report [1] states that the spectra measurements were performed one by one, we conclude therefore that the other experimental channels were closed (plugged with tungsten). The dosimetric foils were irradiated together and simultaneously in plural detector channels.
  
No information on the background correction was found in the literature.
  
Six techniques were employed to measure neutrons and gamma-rays:
  
(1) 14 mm-diameter spherical NE213 liquid organic scintillator was used as the fast (above 2 MeV) neutron spectrometer. It was inserted into the experimental channel hole of 22 mm diameter.
  
(2) A pair of proton recoil gas proportional counters (PRC) was used to measure neutron spectrum in the 5 keV to 1 MeV energy range. The counter has a cylindrical shape with the outer diameter of 19 mm and the effective length of 127 mm. The counter is inserted into an experimental hole of 21 mm in diameter with its pre-amplifier. PRC is made of type 304 stainless steel of a thickness of 0.41 mm.
  
Two types of counters were used to measure neutron spectrum in a wide energy range. Hydrogen gas at 0.5677 MPa (5.789 kgf/cm2) with 1 percent of CH4 filled counter was used for low-energy neutrons from 3 keV to 150 keV while for the high-energy neutrons from 150 keV to 1 MeV the counter was filled with 50-50 mixture of hydrogen and argon gases with 1.8 percent of nitrogen at 0.6102 MPa (6.222 kgf/cm2).
  
Error Assessment: Possible error sources are gas pressure (number of hydrogen atom), n-p scattering cross section, fitting error for differentiation of recoil proton spectrum due to count statistics and calibration of recoil proton energy. The fitting error is the largest, ~ 3-10 % above 10 keV, while the other errors are expected to be less than 1%. Neutron spectra below 10 keV tend to become smaller due to the uncertainty of the W-value, which is the average energy loss per ion pair. The error due to W-value is not included in the experimental errors.
  
(3) Slowing down time (SDT) method for the neutron spectrum from 1 eV to 300 eV. A BF3 gas proportional counter with an outer diameter of 14 mm and an effective length of 99 mm, containing 96 % boron-10 (B-10) enriched BF3 gas at the pressure of 71.5 kPa, was used for neutron detection. Using the standard thermal neutron field the effective number of B-10 atoms in the counter was determined to be 2.18E20 +- 3 %. The counter was inserted into one of the experimental holes of the experimental assembly.
  
(4) Dosimetry reaction rate of the Al-27(n,alpha)Na-24, Nb-93(n,2n)Nb-92m, In-115(n,n')In-115m, 186W(n,gamma)187W and Au-197(n,gamma)Au-198 reactions were measured by the foil activation method. The characteristics for the reactions used are provided. Typical sample size was 10 mm in diameter and 1 mm in thickness for activation foils except indium and gold. Indium foils had dimensions of 10 x 10 x 1 mm3. In order to minimize the self-shielding effect for the Au-197(n,gamma) reaction, gold foils with a size of 10 x 10 x 0.001 mm3 were adopted.
  
Experimental Error and Uncertainty: Major sources of the error for the reaction rate were the gamma-ray counting statistics (0.1 ~ several %) and the detector efficiency (2 ~ 3 %). The error for sum-peak correction was estimated less than 2 % depending on the decay mode and fraction of multiple gamma-ray cascade. The error for the decay correction was reflected from the error of half-life of the activity. If the half-life was accurate, the error for the saturation factor should be less than 1 % even for the short half-life activities. The other errors associated with foil weight, gamma-ray self-absorption, irradiation time, cooling time and counting time were negligibly small. The error for neutron yield was estimated to be 2 %.  The overall error for the major part of reaction rate ranged between 3 ~ 6 %.  Some data for high threshold reaction in the deep positions suffered from poor counting statistics due to low activation rate.
  
(5) Prompt gamma-ray spectrum were measured by a 40 mm diameter spherical BC537 liquid organic scintillation counter. Outer diameter of the detector is 48 mm and length including a photomultiplier assembly is 262 mm. Experimental uncertainty: As explained in (6) below, typical experimental uncertainties of the gamma-ray heating rate measured by TLDs range between 7 ~ 15 %. The gamma-ray spectra are normalized to the gamma-ray heating rates. Uncertainties of about 10 % is introduced by the normalization procedure. All of the rest of experimental uncertainties, such as statistical errors, uncertainties of the response functions, determination of source D-T neutrons and subtraction of decay gamma-rays, are less than 5 %.  Therefore, total experimental uncertainties are approximately 15 ~ 20 %. Note that only the statistical erorrs are included.
  
(6) Gamma-ray heating rate were measured by thermoluminescent dosimeters (TLDs) in combination with the atomic number interpolation method. Gamma-ray heating rates of vanadium were deduced by interpolating the gamma-ray heating rates measured by three types of TLDs, Mg2SiO4 (MSO, effective atomic number Zeff = 11.1), Sr2SiO4 (SSO, Zeff = 32.5) and Ba2SiO4 (BSO, Zeff = 49.9).
  
Sources of error in the measured gamma-ray heating rates are as follows:
  
Statistical deviation of four TLDs      5 - 15 %
Number of neutrons generated            2 - 3  %
Calibration of the TLD reader             5    %
  
Due to the subtraction of target gamma-rays and neutron response, the following uncertainties are to be added to the above errors according to the quadratic propagation of error.
  
Response functions for neutrons    30 %
Neutron energy spectra             10 %
Target gamma-ray                   20 %
  
Overall errors for the obtained gamma-ray heating rates of vanadium are ~ 10%, except at the positions in the front surface of the assemblies where they are 20 ~ 25 %.
  
Description of Results and Analysis:
-----------------------------------
The measured neutron spectra by the NE213 and PRC at the depths of 76 mm, 228 mm and 380 mm are shown. The numerical data of the spectra measured by the NE213 and PRC are given. A measurement of neutron spectra below 10 keV by the SDT method was attempted. However, it could not be made because of too low neutron flux in the energy range. The measured dosimetry reaction rates and the measured gamma-ray spectra at the three positions in the tungsten assembly are shown. The numerical data of the spectra are given. The measured gamma-ray heating rates in tungsten are given. Most of the observed gamma-rays at the front surface of the assembly, 0 mm, are produced by high energy neutron of > 1 MeV. At 76 mm, both halves of observed gamma-rays are produced by neutrons of > 1 MeV and 0.01-1 MeV. At 228 and 380 mm, neutrons with energy between 1 keV - 1 MeV contribute predominantly to produce secondary gamma-rays.
  
Transport calculations with the MCNP-4A code were performed using the JENDL-3.2, JENDL-FF and FENDL/E-1.0 nuclear data libraries.
  
The calculations were performed at the OECD/NEA Data Bank [5] using the included MCNP inputs and the FENDL-2 (=JENDL-FF) and ENDF/B-VI.8 cross section evaluations. The neutron spectra and the corresponding C/E ratios are presented. The measured fusion peak is wider than the one calculated with the provided MCNP input. The gamma spectra obtained using the ENDF/B-VI.3 data indicate reasonable agreement between the experiment and the calculation.

NEA-1553/75
SINBAD-IPPE-FE
==============
Measurement System and Uncertainties:
------------------------------------
The detector used was a fast scintillator detector located at an angle of 8 deg. relative to beam trajectory extension and at a flight path of 6.8 m.
The detector was installed in a lead house behind a concrete wall. A conical hole drilled through the wall acted as a collimator.
The detector itself consisted of a cylindrical paraterphenyl cristal of 5 cm diameter and 5 cm height. It was coupled to a FEU-143 photomultiplier.
The time-of-flight measurement is made in the usual inverse method, i.e. using the detector signal as a start signal and the delayed deuteron pick-up pulse as a stop signal. In this way only the useful neutron bursts, i.e. those producing a signal in the detector are used, so avoiding dead time losses.
To experimental spectra were corrected for the background effects. To measure the background neutron spectra, a 1 m long by 18 - 26 cm diameter iron shadow bar and a 30 cm long borated polyethylene cylinder were placed between the detector and the sphere.
The estimated uncertainties of the experimental data and their main components are listed. During the experiment the main spectrometer parameters (detector efficiency, absolute normalization factor, etc.) were measured several times, hence the stability of the spectrometer could be estimated by calculating the mean square deviation of individual runs.
Two radioactive reference sources were used, Cf252 for neutron detector calibration and Pu238 for alfa detector calibration, with their specific uncertainties. The uncertainties of corrections for Cf-chamber scattering and time-of-flight conversion to energy, calculated with MCNP, were estimated at about 1-2%.
The quadratic sum of components 2-5, considered as systematic, is calculated and its quadratic summation with the statistical uncertainty gives the total uncertainty of the experimental data.
  
Description of Results and Analysis:
-----------------------------------
The measured TOF spectra were corrected for the background effects and converted to the energy spectrum. The leakage spectrum, L(E), representing the differential fluence of leakage neutrons, integrated over the full sphere (4 pi sr) and normalised to 1 source neutron, was then calculated from the following expression:

L(E)=4*pi*N(E)/(eps(E)*dOmega*Nn)

where:
N(E)   = neutron energy spectrum, converted from measured TOF spectra,
eps(E) = neutron detection efficiency,
dOmega = detector solid angle (=(pi*r*r)/(L*L), where r is the detector radius and L the distance from the sphere to the detector), Nn   = number of source neutrons.

The experimental results are presented in
Table 4 for iron shell No. 1 (r= 4.5 cm, wall thickness = 2.5 cm),
Table 5 for iron shell No. 2 (r=12.0 cm, wall thickness = 7.5 cm),
Table 6 for iron shell No. 3 (r=12.0 cm, wall thickness =10.0 cm),
Table 7 for iron shell No. 4 (r=20.0 cm, wall thickness =18.1 cm),
Table 8 for iron shell No. 5 (r=30.0 cm, wall thickness =28.0 cm)

as leakage spectrum in terms of neutrons per MeV and per source neutron.

MCNP-4C input data for Shells #1 - #5 (4.5 - 30 cm radius) are given in files mcnp_fe1.inp, mcnp_fe2.inp, mcnp_fe3.inp, mcnp_fe4.inp and mcnp_fe5.inp.
In the models the spheres and the neutron source are described pricisely, including anisotropic energy and yield distributions of the T(d,n) source.
For an adequate comparison of measurements and analitical calculation, the convolution with the spectrometer response function, describing the energy resolution of the spectrometer, is necessary.
A correction factor was applied for the experimental spectra to account for the time-of-flight procedure with bulk smaples. This has been done with the
Monte Carlo time-independent calculations.
Refs. [1] and  [4] discuss also the corrections for non-spherical effects,
which should be taken into account in case of 1-dimensional (spherical) calculations using codes like ANISN, ONEDANT, ANTRA-1 etc.

NEA-1553/76
SINBAD-IPPE-V
=============
Measurement System and Uncertainties:
-------------------------------------
The detector used was a fast scintillator detector located at an angle of 8 deg. relative to beam trajectory extension and at a flight path of 6.8 m. The detector was installed in a lead house behind a concrete wall. A conical hole drilled through the wall acted as a collimator.
  
The detector itself consisted of a cylindrical paraterphenyl cristal of 5 cm diameter and 5 cm height. It was coupled to a FEU-143 photomultiplier.  
  
The time-of-flight measurement is made in the usual inverse method, i.e. using the detector signal as a start signal and the delayed neutron source signal as a stop signal. In this way only the useful neutron bursts, i.e. those producing a signal in the detector are used, so avoiding dead time losses.  
  
The experimental spectra were corrected for the background effects. To measure the background neutron spectra, a 1 m long by 18 - 26 cm diameter iron shadow bar and a 30 cm long borated polyethylene cylinder were placed between the detector and the sphere.  
  
The estimated uncertainties of the experimental data and their main components are listed. Their dependence on leakage neutron energy in the case of the smaller shell is shown. During the experiment the main spectrometer parameters (detector efficiency, absolute normalization factor, etc.) were measured several times, hence the stability of the spectrometer could be estimated by calculating the mean square deviation of individual runs.
  
Two radioactive reference sources were used, Cf252 for neutron detector calibration and Pu238 for alpha detector calibration, with their specific uncertainties. The uncertainties of corrections for Cf-chamber scattering and time-of-flight conversion to energy, calculated with MCNP, were estimated at about 1-2%.  
  
The quadratic sum of components 2-5, considered as systematic, is calculated and its quadratic summation with thestatistical uncertainty gives the total uncertainty of the experimental data.
  
Description of Results and Analysis:
------------------------------------
The measured TOF spectra were corrected for the background effects and converted into the energy spectrum. The leakage spectrum, L(E), representing the differential fluence of leakage neutrons, integrated over the full sphere (4 pi sr) and normalised to 1 source neutron, was then calculated from the following expression:
  
    L(E)=4*pi*N(E)/(eps(E)*dOmega*Nn)
  
where:
N(E)  = neutron energy spectrum, converted from measured TOF spectra,
eps(E) = neutron detection efficiency (see Ref. [2] for details),
dOmega = detector solid angle (=(pi*r*r)/(L*L), where r is the detector radius and L the distance from the sphere to the detector),
Nn = number of source neutrons.
  
Are presented: the results for vanadium shell No. 1 as leakage spectrum in terms of neutrons per MeV and per source neutron are presented, and for vanadium shell No. 2 the measured leakage spectrum.
  
MCNP-4C input data for Shells 1 and 2 (5 & 12 cm radius) are given. In the models the spheres and the neutron source are described precisely, including anisotropic energy and yield distributions of the T(d,n) source.
  
For an adequate comparison of measurements and analytical calculation, the convolution with the spectrometer response function, describing the energy resolution of the spectrometer, is necessary.
  
A correction factor should also be applied before comparing the experimental spectra with the Monte Carlo time-independent calculations. The function that should be multiplied with the experimental spectrum is shown.
   
The 2 references discuss also the corrections for non-spherical effects, which should be taken into account in case of 1-dimensional (spherical) calculations using codes like ANISN, ONEDANT, ANTRA-1 etc.

NEA-1553/77
SINBAD-OKTAVIAN/AL
==================
Measurement System:
------------------
A cylindrical liquid organic scintillator NE-218 (12.7 cm-diam, 5.1 cm-long) was used as a neutron detector. The detector efficiency was determined by combining:
1) the Monte Carlo calculation,
2) the measured efficiency derived from the TOF measurement of Cf-252 spontaneous fission spectrum and the Watt's spectrum, and
3) the measured efficiency from the leakage spectrum from a graphite sphere, 30 cm in diameter with the similar detection system.
  
To monitor the absolute neutron spectrum per source neutron, a cylindrical niobium foil was set in front of the tritium target and irradiated during the TOF experiment. From the gamma-ray intensity of the induced activity, Nb-92m and the integrated counts of the source neutron spectrum, the absolute neutron leakage spectrum can be obtained. The formulation of this procedure is described in the Oktavian Report [4].
  
To measure the gamma spectra, OKTAVIAN was run in the pulsed mode with a repetition frequency of 500 kHz. The pulse width was 3 ns in FWHM and the difference in flight times between the 14 MeV neutrons and the prompt gamma-rays was about 90 ns from the sphere to the detector. Since those were enough to separate the gamma-rays from the neutron background in the TOF spectra, the desired gamma-rays could be discriminated from a neutron background.
  
The gamma emission spectra were dominated by the gamma-rays from (n,n') and (n,2n) reactions rather than the gamma-rays from (n,xgamma) reaction. The data are therefore available to the assessment in the nuclear data for energy distributions of gamma-rays from non-elastic scattering by high energy neutrons.
  
Description of Results and Analysis:
-----------------------------------
Source of Information:
The main source of information were references [5] and [6]. The information on the source neutron spectrum is ambuguous. Namely, the spectrum in the text is given on a different energy grid than the spectrum in the sample MCNP input in ref. [5] on pages 80 and 124, respectively. Furthermore, there is a trivial error in the exponent in the spectrum at about 0.5 MeV in the sample MCNP input, which is also evident as an unusual bump in the calculated neutron leakage spectrum below 0.5 MeV in Fig. 4.8 of the same document. The same error persist even in a more recent document [7]. By contacting the author it has been established that the recommended source spectrum for the calculation is the one from the sample MCNP input, corrected for the trivial error in the exponent. The energy grid in this spectrum is more refined around the 14 MeV peak and hence better suited for the calculations.
  
Error Assessment:
The experimental errors in the measured neutron spectra include only statistical deviation (1 sigma). The relative error to measure the niobium activation foils is less than 1 % (0.4 to 1 %), which is not included.
  
In the measured gamma spectra the following sources were included in the errors:
(a) Uncertainty in monitoring absolute fluxes of the source neutrons
(b) Errors of the response matrix
(c) Statistical deviation (lcs)
  
Example of Experiment Analysis:
Three sets of inputs are provided
- two older 1-dimensional (1D) MCNP models for any 55deg. line experiment. One model (file mcnp4b.inp) includes neutron, photon and electron transport with the neutron source term. The other (mcnp4b_g.inp) includes photon and electron transport with the gamma source term.
- two routine MCNPX(5) (semi) 2-D models in which the neutron source or the gamma source is specified (AL2dns.i, AL2dgs.i).
- Detailed 3-dimensional MCNPX(5) model including the full experimental information for both neutron and gamma spectra (AL3dn.i).
- Detailed 3-dimensional MCNPX(5) model including the full experimental information and explicit D-T source for gamma spectra analysis. (AL3dg.i). The D-T source routine (patch_DT) is needed to run this input.
  
The results obtained using the (obsolete) 1-D model and different cross-section evaluations (ENDF/B-VI.1, EFF-3) are compared in Figure 3 for neutron spectrum below 1 MeV and in Figure 4 for the high energy part of the neutron spectrum. Full energy range comparison is shown in Figure 5. Similarly, the gamma spectrum measurement results are compared in Figure 6, emphasizing the low energy spectrum and in Figure 7 for the high energy part of the spectrum.
  
Results using more accurate 2-D and 3-D models are discussed in [9].

NEA-1553/78
SINBAD-OKTAVIAN-MN
=================
Measurement System:
------------------
A cylindrical liquid organic scintillator NE-218 (12.7 cm-diameter, 5.1 cm-long) was used as a neutron detector. The detector efficiency was determined by combining:
1) Monte Carlo calculation,
2) measured efficiency derived from the TOF measurement of Cf-252 spontaneous fission spectrum and the Watt's spectrum, and
3) measured efficiency from the leakage spectrum from a graphite sphere, 30 cm in diameter with the similar detection system.

To monitor the absolute neutron spectrum per source neutron, a cylindrical niobium foil was set in front of the tritium target and irradiated during the TOF experiment. From the gamma-ray intensity of the induced activity, Nb-92m and the integrated counts of the source neutron spectrum, the absolute neutron leakage spectrum can be obtained. The formulation of this procedure is described in the Oktavian Report [5].

To measure the gamma spectra, OKTAVIAN was run in the pulsed mode with a repetition frequency of 500 kHz. The pulse width was 3 ns in FWHM and the difference in flight times between the 14 MeV neutrons and the prompt gamma-rays was about 90 ns from the sphere to the detector. Since those were enough to separate the gamma-rays from the neutron background in the TOF spectra, the desired gamma-rays could be discriminated from a neutron background.
  
Description of Results and Analysis:
-----------------------------------
The measured neutron leakage spectrum from a 60 cm diameter Manganese pile is given in Table 3. The numerical data for the experimental gamma-ray leakage spectra, measured from teh 60 cm diameter Manganese, pile is given in Table 4. Error Assessment: The experimental errors in the measured neutron spectra include only statistical deviation (1 s). The relative error to measure the niobium activation foils is less than 1 % (0.4 to 1 %), which is not included. In the measured gamma spectra the following sources were included in the errors:
(a) Uncertainty in monitoring absolute fluxes of the source neutrons,
(b) Errors of the response matrix,
(c) Statistical deviation (1 s).

Analysis:
Two MCNP5(X) models are provided:
- Mn2dn.i is a simple routine model with neutron source. The neutron source is defined according to the OKTAVIAN anisotropy specifications. At any angle corresponds the neutron yielding and energy (discrete values). According to methods discussed in [4], the neutron source energy spectrum in Table 1 may be converted into time domain to perform the resolution broadening of the time-of-flight spectrum calculated with routine model 2dmn.i. Then, the folded spectrum may be converted from time into energy domain and compared with experimental neutron leakage spectrum (Table 3). This model can easily be modified to calculate the gamma leakage spectrum with neutron source.
-  Mn3dn.i is a detailed 3?dimensional  model including the full experimental (concerning geometry and material composition) information for neutron spectra calculations. The source term in the model corresponds to the OKTAVIAN discrete neutron source specifications. The resolution broadening has to be applied afterwards [4].
top ]
7. UNUSUAL FEATURES: SPECIAL FEATURES
NEA-1553/27
SINBAD-SB5-FUS
==============
Relatively high-energy neutron source. Significant data base of gamma-ray measurements.

NEA-1553/41
SINBAD-KANT
===========
Other than in most experiments on fusion neutron transport, the leakage neutron spectra were measured through the full energy range from the 14 MeV peak down to thermal energy.

NEA-1553/46
SINBAD-TUD-FE
=============
Quality assessment:
------------------
The TUD IRON SLAB experiment is ranked as benchmark quality experiment. For detailed evaluation see document IJS-DP-10216 (April 2009) by A. Milocco.

NEA-1553/47
SINBAD-TUD-FNG-W
============
Quality assessment:
------------------
The TUD/FNG TUNGSTEN experiment could be ranked as benchmark quality experiment, provided that supplementary experimental information is available on:- realistic and complete estimation of neutron and gamma flux point-wise uncertainties,
- availability of the original pulse-height distributions measured by spectrometers and the detector geometry would be useful for those who wish to carry out their own spectra unfolding,
- some inconsistencies observed with the FNG-W benchmark results should be explained and resolved.
For detailed evaluation see Ref. [8].

NEA-1553/48
SINBAD-OKTAVIAN/NI
==================
Quality assessment:
------------------
The OKTAVIAN NICKEL experiment is of benchmark quality for nuclear data validation purposes. For detailed evaluation see "The Quality Assessment of the OKTAVIAN Benchmark Experiments, IJS-DP-10214, April 2009".

NEA-1553/50
SINBAD-OKTAVIAN/FE
==================
Quality assessment:
-----------------
The OKTAVIAN IRON experiment seems to be of sufficient quality for nuclear data validation purposes. However, the measurements above 4 MeV should be used with caution.
For detailed evaluation see document "The Quality Assessment of the OKTAVIAN Benchmark Experiments", IJS-DP-10214, April 2009.

NEA-1553/52
SINBAD-OKTAVIAN/SI
==================
Quality assessment:
-----------------
The OKTAVIAN SILICON 60 CM experiment can be ranked as a benchmark quality experiment for nuclear data validation purposes. In order to make a complete use of this benchmark experiment, supplementary experimental information is advisable on:
- the neutron realistic effects below 3 MeV (in particular the background subtraction method should be detailed)
- the gamma source measurements
- the gamma detector response function
- the gamma detector calibration


The OKTAVIAN SILICON 40 CM experiment is ranked as a benchmark experiment of INTERMEDIATE quality because the neutron leakage flux measurements are only available in graphical form and their reading is approximate. Moreover,supplementary experimental information is advisable on:
- the gamma source measurements
- the gamma detector response function
  
For detailed evaluation see [10].

NEA-1553/53
SINBAD-OKTAVIAN-W
=================
Quality assessment:
-----------------
The OKTAVIAN SILICON 40 CM and 60 CM experiments seems to be of sufficient quality for nuclear data validation purposes. However, in order to use this benchmark for the validation of modern cross-section evaluations, supplementary experimental information would be needed on:
- the neutron realistic effects below 0.5 MeV (background subtraction method in particular)
- the gamma source measurements
- the gamma detector response function

For detailed evaluation see [11].

NEA-1553/54
SINBAD-FNG-BLKT
===============
Quality assessment:
------------------
The ITER BLANKET BULK SHIELD experiment is ranked as benchmark quality experiment (for detailed evaluation see [8]).

NEA-1553/55
SINBAD-FNG-DOSE-RATE
====================
Quality assessment:
-----------------
The ITER DOSE RATE experiment is ranked as benchmark quality experiment (for detailed evaluation see [12]).

NEA-1553/56
SINBAD-FNG-SIC
==============
Quality assessment:
------------
The FNG SILICON CARBIDE experiment is ranked as benchmark quality experiment. For detailed evaluation see Ref. [10].

NEA-1553/57
SINBAD-FNG-SS
=============
Quality assessment:
------------------
The STAINLESS STEEL BULK SHIELD experiment is ranked as benchmark quality experiment. However, the fact that the geometrical data are only given in MCNP input format may pose problems for users of other codes. A comprehensive geometry description would be helpful. For detailed evaluation see [3].

NEA-1553/58
SINBAD-FNG-STREAMING
====================
Quality assessment:
-----------------
The ITER NEUTRON STREAMING experiment is ranked as benchmark quality experiment. For detailed evaluation see [16]).

NEA-1553/59
SINBAD-FNG-W
============
Quality assessment:
-----------------
The FNG TUNGSTEN experiment is ranked as benchmark quality experiment (for detailed evaluation see [12]).

NEA-1553/60
SINBAD-FNS-DUCT
===============
Quality assessment:
------------------
The FNS DOGLEG?DUCT experiment is ranked as intermediate quality benchmark experiment. Supplementary experimental information is needed on the neutron source spectrum. Likewise, supplementary experimental information would be useful on:
? the experimental energy calibration
? the neutron detector response function
Note also that the calculated neutron spectra and the reaction rates were normalised so that the 93Nb(n,2n)92mNb reaction rate is consistent with the measured value at the duct inlet.

For detailed evaluation see [9].

NEA-1553/61
SINBAD-FNS-OXYGEN
=================
Quality assessment:
------------------
The FNS OXYGEN experiment is ranked as benchmark quality experiment. However, in order to use this benchmark for the validation of modern cross-section evaluations, supplementary experimental information would be needed on:
? the neutron effective flight path parameter.

For detailed evaluation see [11].

NEA-1553/62
SINBAD-FNS-SKYSHINE
===================
Quality assessment:
------------------
The FNS SKY?SHINE experiment could be ranked as benchmark quality experiment, provided that supplementary information on the neutron source spectrum is supplied.

For detailed evaluation see [3].

NEA-1553/69
SINBAD-TUD-FNG-BS
=================
Quality assessment:
------------
The TUD/FNG ITER Bulk Shield experiment could be ranked as benchmark quality experiment, provided that supplementary experimental information is available:
? realistic and complete estimation of neutron and gamma flux point?wise uncertainties;
- availability of the original pulse-height distributions measured by spectrometers and the detector geometry would be useful for those who wish to carry out their own spectra unfolding.

For detailed evaluation see Ref. [15].

NEA-1553/70
SINBAD-TUD-FNG-SIC
==================
Quality assessment:
------------
The TUD/FNG SILICON CARBIDE experiment could be ranked as benchmark quality experiment, provided that supplementary experimental information is available on:
? realistic and complete estimation of neutron and gamma flux point?wise uncertainties,
- availability of the original pulse-height distributions measured by spectrometers and the detector geometry would be useful for those who wish to carry out their own spectra unfolding,
- some inconsistencies observed with the FNG-W benchmark results should be explained and resolved.

For detailed evaluation see Ref. [11].

NEA-1553/71
SINBAD-FNG-HCPB
===============
Quality assessment:
------------------
The FNG HCLL experiment is suitable for benchmark purposes.

NEA-1553/72
SINBAD-FNS-C-CYLIND
===================
Quality assessment:
-------------------
The FNS GRAPHITE experiment is suitable for benchmark purposes.
However, in order to use this benchmark for the validation of modern cross-
section evaluations, supplementary experimental information would be needed on:
? effect of the experimental unfolding technique
? activation foils arrangement
- TLD measurements

For detailed evaluation see [11].

NEA-1553/73
SINBAD-FNS-W
================
Quality assessment:
------------------
The FNS VANADIUM experiment is is suitable for benchmark purposes.
However, in order to use this benchmark for the validation of modern cross-section evaluations, supplementary experimental information would be needed on:
? effect of the experimental unfolding technique of the Ne?213 measurements
? activation foils positioning, corresponding uncertainty and housing

For detailed evaluation see [5].

NEA-1553/74
SINBAD-FNS-W
================
Quality assessment:
------------------
The FNS TUNGSTEN experiment is suitable for benchmark purposes.
However, in order to use this benchmark for the validation of modern cross-section evaluations, supplementary experimental information would be needed on:
? effect of the experimental unfolding technique of the Ne?213 measurements
- activation foils positioning, corresponding uncertainty and housing

For detailed evaluation see [6].

NEA-1553/75
SINBAD-IPPE-FE
==============
Quality assessment:
----------------
In order to use the IPPE Iron experiment for the validation of modern cross-section evaluations, supplementary experimental information would be needed on the experimental set-up. In particular, the specifications for the collimator duct are of major concern for the calculation of the leakage spectra in the energy region of the resolved resonances.

For detailed evaluation see [5].

NEA-1553/76
SINBAD-IPPE-V
===========
Quality assessment:
-----------------
In order to use the IPPE Vanadium experiment for the validation of modern cross-section evaluations, supplementary experimental information would be needed on the experimental set-up. In particular, the specifications for the collimator duct are of major concern for the calculation of the leakage spectra in the energy region of the resolved resonances.

For detailed evaluation see [4].

NEA-1553/77
SINBAD-OKTAVIAN/AL
==================
Quality assessment:
------------------
The OKTAVIAN ALUMINIUM experiment is suitable for benchmark purposes.
However, in order to use this benchmark for the validation of modern cross-
section evaluations, supplementary experimental information is advisable on:
? the neutron flight path parameter
? the neutron realistic effects below 1 MeV (especially background subtraction method)
? the gamma source measurements
? the gamma detector response function.

For detailed evaluation see [9].

NEA-1553/78
SINBAD-OKTAVIAN-
================
Quality assessment:
------------------
The OKTAVIAN MANGANESE experiment seems to be of sufficient quality for nuclear data validation purposes.
However, in order to use this benchmark for the validation of modern cross-section evaluations, supplementary experimental information would be needed on:
? the neutron realistic effects, in particular those concerning the background subtraction method
? the gamma source measurements

For detailed evaluation in similar experiments see [4].
top ]
9. STATUS
Package ID Status date Status
NEA-1553/01 02-JUN-1996 Tested at NEADB
NEA-1553/26 01-DEC-2000 Masterfiled Arrived
NEA-1553/27 01-DEC-2000 Masterfiled Arrived
NEA-1553/40 15-DEC-2005 Tested at NEADB
NEA-1553/41 13-MAR-2006 Tested at NEADB
NEA-1553/43 04-DEC-2008 Masterfiled Arrived
NEA-1553/45 06-FEB-2009 Masterfiled Arrived
NEA-1553/46 07-MAY-2010 Tested at NEADB
NEA-1553/47 05-MAY-2010 Tested at NEADB
NEA-1553/48 03-MAY-2010 Masterfiled Arrived
NEA-1553/50 05-MAY-2010 Tested at NEADB
NEA-1553/52 05-MAY-2010 Tested at NEADB
NEA-1553/53 07-MAY-2010 Tested at NEADB
NEA-1553/54 20-DEC-2011 Masterfiled Arrived
NEA-1553/55 20-DEC-2011 Masterfiled Arrived
NEA-1553/56 20-DEC-2011 Masterfiled Arrived
NEA-1553/57 20-DEC-2011 Masterfiled Arrived
NEA-1553/58 20-DEC-2011 Masterfiled Arrived
NEA-1553/59 20-DEC-2011 Masterfiled Arrived
NEA-1553/60 20-DEC-2011 Masterfiled Arrived
NEA-1553/61 20-DEC-2011 Masterfiled Arrived
NEA-1553/62 20-DEC-2011 Masterfiled Arrived
NEA-1553/69 21-DEC-2011 Masterfiled Arrived
NEA-1553/70 21-DEC-2011 Masterfiled Arrived
NEA-1553/71 14-MAR-2012 Masterfiled Arrived
NEA-1553/72 14-MAR-2012 Masterfiled Arrived
NEA-1553/73 01-MAR-2012 Masterfiled Arrived
NEA-1553/74 01-MAR-2012 Masterfiled Arrived
NEA-1553/75 01-MAR-2012 Masterfiled Arrived
NEA-1553/76 01-MAR-2012 Masterfiled Arrived
NEA-1553/77 01-MAR-2012 Masterfiled Arrived
NEA-1553/78 01-MAR-2012 Masterfiled Arrived
top ]
10. REFERENCES

- Ivo Kodeli, Hamilton Hunter, Enrico Sartori:
Radiation Shielding and Dosimetry Experiments Updates in the SINBAD database
Radiation Protection Dosimetry (2005), Vol.116, No.1-4, pp.558-561
NEA-1553/26, bibliography:
SINBAD-ILL-FE
=============
Background references:
[1] R.H. Johnson, "Integral Tests of Neutron Cross Sections for Iron, Nobium, Beryllium, and Polyethylene," PhD Thesis, University of Illinois at Urbana-Champaign (1975)
[3] M.L. Williams, C. Aboughantous, M. Asgari, J.E. White, R.Q. Wright and F.B.K. Kam, "Transport Calculations of Neutron Transmission Through Steel Using ENDF/B-V, Revised ENDF/B-V, and ENDF/B-VI Iron Evaluations," Annual Nuclear Energy, 18, 549-565 (1991)
[4] D.T. Ingersoll, "Integral Testing of Neutron Cross Sections Using Simultaneous Neutron and Gamma-Ray Measurements," PhD Thesis, University of Illinois at Urbana-Champaign (1977)
NEA-1553/26, included references:
[2] N.E. Hertel, R.H. Johnson, B.W. Wehring, and J.J. Dorning, "Transmission of
Fast Neutrons Through an Iron Sphere," Fusion Technology, 9, 345-361 (Mar 1986)
[5] N.E. Hertel, "High-Energy Neutron Transport Through Tungsten and Iron," PhD
Thesis, University of Illinois at Urbana-Champaign (1979)
NEA-1553/27, included references:
[1] G. T. Chapman, G. L. Morgan, and J. W. McConnell, "The ORNL Integral
Experiment to Provide Data for Evaluating Magnetic-Fusion-Energy Shielding
Concepts, Part I: Attenuation Measurements", ORNL/TM/7356 (August 1982).
[2] R. T. Santoro, J. M. Barnes, R. G. Alsmiller, Jr., and E. M. Oblow,
"Calculational Procedures for the Analysis of Integral Experiments for Fusion
Reactor Design", ORNL-5777 (July 1981).Oak Ridge National Laboratory, Oak
Ridge, Tennessee USA 37831
[3]  Philip F. Rose and R. W. Roussin, eds., "SB5: Fusion Reactor Shielding
Benchmark," in Cross Section Evaluation Working Group Benchmark Specifications
Volume II Supplement, Brookhaven National Laboratory Report BNL 19302, Vol. II
(September 1986)(also ENDF-202).

NEA-1553/40, bibliography:
SINBAD-MEPHI
============
Background references:
[1] Fewell T.R., On evaluation of the alpha counting technique for determining 14 MeV neutron yield, NIM, Amsterdam, V.61 (1), pp. 61-71. 1968.
[2] Trikov L.A., Colevatov Yu. I., Volkov V.S., Methods of spectrometer adjustment by means of radionuclide sources of neutrons, Preprint of INPE No. 1730, Obninsk, INPE, 1985 (in Russian).
[3] H.Hashikura et al., Calculation of neutron response of thermoluminiscent dosimeters, J.of Faculty of Eng, University of Tokio, Vol. XXXIX, No1 pp7-16, 1987.
[4] Afanasiev V.V. Andreev M.I., Belevitin A.G., Romodanov V.L.,
Benchmark-Experiments and Analyses on Streaming of 14-MeV Neutrons in Iron and Iron-Water Radiation Shielding Mock-Ups ith Slits, Preprint 003-94, MEPhI, Moscow, 1994.
[5] V.V. Afanasiev, Andreev M.I., Belevitin A.G., Romodanov V.L., at al., Benchmark Experiments and Analyses on Streaming of 14-MeV Neutrons in Iron and Iron-Water Radiation Shielding Mock-Ups with Slits, Topical Meeting, Radiation Protection and Shieldings. No. Falmouth, April 21-25, Vol.2, p. 687. 1996.
[6] Afanasiev V.V., Andreev M.I., Belevitin A.G., Dmitriev D.M., Romodanov V.L., Experimental and calculational studies of non-uniform shieldings of fusion reactors, Proc. ISFNT-4, April 1997, Japan, p.271.
[7] V.V. Afanasiev, Andreev M.I., Belevitin A.G., Romodanov V.L., Dmitriev D.M at al., Benchmark experiments on non-uniform iron shielding compositions (ISTC project No. 180), Proc. ICENES98 - 9th International Conference on Emerging Nuclear Energy Systems, Tel-Aviv, Israel, June 28-July 2, 1998 pp.407-413.
NEA-1553/40, included references:
[8] Afanasiev V.V., Andreev M.I., Belevitin A.G., Dmitriev D.M., Romodanov V.L.:
Experimental and calculational studies of non-uniform shieldings of fusion
reactor, Fusion Engineering and Design, 42, 1998, 261-266.
[9] Andreev M.I., Afanasiev V.V., Belevitin A.G., A.V.Karaulov, Romodanov V.L.
et al.:
Set of benchmark experiments on slit shielding compositions of thermonuclear
reactors, Fusion Engineering and Design, 55, 2001, 373-385.

NEA-1553/41, bibliography:
SINBAD-KANT
===========
Background references:
[B82] G. Bulski et al., Int. Conf. Nuclear Data for Science and Technology, Antwerp, Sept. 1982
[D82] G. Dietze and H. Klein, Report PTB-ND-22 (1982) and unpublished upgrades [E91] H. Ebi, W. Eyrich, H. Fries et al., Fus. Eng. Des. 18 (1991) 317-322
[F00] U. Fischer, R. L. Perel and H. Tsige-Tamirat, Fus. Eng. Des. 51-52 (2000) 761-768
[H94] K. Hayashi, U.v.Moellendorff, T. Tsukiyama et al., 'Fusion Technology 1994' (Eds. K. Herschbach et al.), Elsevier (1995), Vol. 2, 1349-1352
[Ka73] F. Kappler, D. Rusch and E. Wattecamps, Nucl. Instr. Meth. 111 (1973) 83-92
[Ke73] C.D. Kemshall, Report AWRE 031 (1973)
[M95] U. von Moellendorff, A.V. Alevra, H. Giese et al., Fus. Eng. Des. 28 (1995) 737-744
[T01] R. Tayama, T. Tsukiyama, K. Hayashi et al., Fus. Eng. Des. 55 (2001) 365-372
[Y79] P. G. Young and L. Stewart, Report LA-7932-MS, 1979
NEA-1553/41, included references:
[M95] U. von Moellendorff, A.V. Alevra, H. Giese, F. Kappler, H. Klein and R.
Tayama: Measurements of 14 MeV Neutron Multiplication in Spherical Beryllium
Shells Fusion Engineering and Design 28(1995)737-744

NEA-1553/43, bibliography:
SINBAD-LI-BLANKET
=================
Background references:
[6] R. Dierckx, Nucl. Instr.& Meth. 107,397  (1973)
[11] R. Hecker, P. Cloth, D. Filges, "Survey on Experimental Neutron Physics of CTR Blankets in the KFA", Proceedings of the 9th Symposium on Fusion Technology 1976, EUR-5602, pp.551-556; CONF-76063
[12] P. Cloth, D. Filges R. Herzing, N. Kirch, "Neutron Multiplication Effect of CTR Blankets Containing Beryllium", Proceedings of the 9th Symposium on Fusion Technology 1976, EUR-5602, pp.569-575; CONF-760631
NEA-1553/43, included references:
[1] L. Kuijpers, "Experimental Model Studies for a Fusion Reactor  Blanket",
KFA Juelich Report 1356(1976)
[2] L. Herzing,  " Erprobung neutronenphysikalischer Rechenverfahren on
Lithiumblanketmodellen fuer einen Fusionsreaktor",  KFA Juelich Report
1357(1976)
[3] G. Gyorgen, KFA Juelich Report 2060(1986).
[4] W. Pohorecki, "Diagnostics of some parameters of a fusion reactor blanket
model",  AGH MIFiTJ Report INT 239/PS ISSN 0302-9034, Cracow 1989 (in polish).
[5] S. M. Quaim, R. Woelfle, G. Stoecklin, "Radiochemical Methods in the
determination of Nuclear data for Fusion Reactor Technology", Journal of
Radioanalytical Chemistry, Vol. 30 (1976) 35-51
[7] G. Gyorgen, W. Pohorecki, J. M. Zazula, "Calculation of the LiF TLD's Kerma
Factors for Estimation of their Neutron Responses in a Lithium Blanket Model",
KFA-IRE-IB-20/84
[8] P. Cloth, D. Filges, K. H. Hammelmann and N. Kirch, "A Homogenious
Lithium-Metal Cylinder for CTR-Blanket Experiments", Nuclear Instruments and
Methods, 124 (1975) 305-306
[9] R. Herzing, L. Kuijpers, P. Cloth, D. Filges, R. Hecker and N. Kirch,
"Experimental and Theoretical Investigations of Tritium Production in a
Controlled Thermonuclear Reactor Blanket Model", Nuc. Sci. Eng., 60, 169-175
(1976)
[10] R. Herzing, L. Kuijpers, P. Cloth, D. Filges, R. Hecker and N. Kirch, "The
Tritium Production in a Controlled Thermonuclear Reactor Blanket Model with a
Graphite Reflector", Nuc. Sci. Eng., 63, 341-343 (1978)
NEA-1553/45, included references:
SINBAD-FNS-BENCHM
=================
- (Ed.) Sub Working Group of Fusion Reactor Physics Subcommittee:
Collection of Experimental Data for Fusion Neutronics Benchmark
JAERI-M 94-014 (February 1994)
- Y. Oyama et al.:
Phase III Experiments of the JAERI/USDOE Collaborative Program on
Fusion Blanket Neutronics - Line Source and Annular Blanket
Experiments - Volume I : Experiment, JAERI-M 94-015 (February 1994)
- F. Maekawa et al.:
Benchmarck Experiment on a Copper Slab Assembly Bombarded by D-T Neutrons
JAERI-M 94-038 (March 1994)

NEA-1553/46, bibliography:
SINBAD-TUD-FE
=============
Background references:
[1] H. Freiesleben, W. Hansen, H. Klein, T. Novotny, D. Richter, R. Schwierz, K. Seidel, M. Tichy, S. Unholzer, Experimental results of an iron slab benchmark, Report Technische Universitaet Dresden, TUD-PHY-94/2, February 1995
[2] H. Freiesleben, W. Hansen, D. Richter, K. Seidel, S. Unholzer, Experimental investigation of neutron and photon penetration and streaming through iron assemblies, Fusion Engineering and Design 28 (1995) 545-550
[3] H. Freiesleben, W. Hansen, D. Richter, K. Seidel, S. Unholzer, Shield Penetration Experiments, Report Technische Universitaet Dresden, Institut fuer Kern- und Teilchenphysik, TUD-IKTP-95/01, January 1995
[4] H. Freiesleben, W. Hansen, D. Richter, K. Seidel, S. Unholzer, TUD experimental benchmarks of Fe nuclear data, Fusion Engineering and Design 37 (1997) 31-37
[5] U. Fischer, H. Freiesleben, H. Klein, W. Mannhardt, D. Richter, D. Schmidt, K. Seidel, S. Tagesen, H. Tsige-Tamirat, S. Unholzer, H. Vonach, Y. Wu, Application of improved neutron cross-section data for Fe-56 to an integral fusion neutronics experiment, Int. Conf. on Nuclear Data for Science and Technology, Trieste (Italy), May 19-24, 1997
[6] M. Tichy, The DIFBAS Program - Description and User's Guide, Report PTB-7.2- 193-1, Braunschweig 1993
[7] S. Guldbakke, H. Klein, A. Meister, J. Pulpan, U. Scheler, M. Tichy, S. Unholzer, Response Matrices of NE213 Scintillation Detectors for Neutrons, Reactor Dosimetry ASTM STP 1228, Ed. H. Farrar et al., American Society for Testing Materials, Philadelphia, 1995, p. 310-322
[8] L. Buermann, S. Ding, S. Guldbakke, S. Klein, H. Novotny, M. Tichy, Response of NE213 Liquid Scintillation Detectors to High-Energy Photons, Nucl. Instr. Methods A 332(1993)483
[9] J. F. Briesmeister (Ed.), MCNP - A general Monte Carlo n-particle transport code, version 4A, Report, Los Alamos National Laboratory, LA-12625-M, November 1993
[10] S. Ganesan and P. K. McLaughlin, FENDL/E - evaluated nuclear data library of neutron interaction cross-sections and photon production cross-sections and photon-atom interaction cross-sections for fusion applications, version 1.0, Report IAEA-NDS-128, Vienna, May 1994
[11] J. Kopecky, H. Gruppelaar, H.A.J. Vanderkamp and D. Nierop, European Fusion File, Version-2, EFF-2, Final report on basic data files, Report, ECN-C-92-036, Petten, June 1992.
[12] Y. Wu, Report FZKA-5953, Karlsruhe, 1997
NEA-1553/46, included references:
[13] A. Milocco:
The Quality Assessment of the FNG/TUD Benchmark Experiments, IJS-DP-10216,
April 2009

NEA-1553/47, bibliography:
SINBAD-TUD-FNG-W
============
Background references:
[1] M. Angelone, M. Pillon, P. Batistoni, M. Martini, M. Martone, V. Rado, "Absolute experimental and numerical calibration of the 14 MeV neutron source at the Frascati Neutron Generator", Rev. Sci. Instr. 67(1996)2189.
[2] M. Tichy, "The DIFBAS Program - Description and User's Guide", Report PTB-7.2- 193-1, Braunschweig 1993.
[3] S. Guldbakke, H. Klein, A. Meister, J. Pulpan, U. Scheler, M. Tichy, S. Unholzer, "Response Matrices of NE213 Scintillation Detectors for Neutrons", Reactor Dosimetry ASTM STP 1228, Ed. H. Farrar et al., American Society for Testing Materials, Philadelphia, 1995, p. 310.
[4] J. F. Briesmeister (Ed.), "MCNP - A general Monte Carlo n-particle transport code", version 4C, Report LA-13709, Los Alamos National Laboratory, 2000.
[5] H. Wienke, M. Herman, "FENDL/MG-2.0 and FENDL/MC-2.0 - The processed cross section libraries for neutron and photon transport calculations", Report IAEA-NDS-128, Vienna, 1998.
NEA-1553/47, included references:
[6] H. Freiesleben, C. Negoita, K. Seidel, S. Unholzer, U. Fischer,
D. Leichtle, M. Angelone, P. Batistoni, M. Pillon:
Measurement and analysis of neutron and gamma-ray flux spectra in Tungsten
Report TUD-IKTP/01-03, Dresden, EFFDOC-857 (2003)
[7] U. Fischer et al.:
Monte Carlo Transport and Sensitivity Analyses for the TUD Neutron
Transport Benchmark Experiment on Tungsten, EFFDOC-860 (2002)
[8] A. Milocco, The Quality Assessment of the FNG/TUD Benchmark Experiments,
IJS-DP-10216, April 2009
[9] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D-T Neutron Source
in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/48, bibliography:
SINBAD-OKTAVIAN/NI
==================
Background references:
[4] A. Takahashi: "Integral Neutronics Experiments at OKTAVIAN", OKTAVIAN Report B-83-01 (1983)
[5] A. Trkov: "Comments on the Oktavian Nickel Sphere Benchmark", Institute Jozef Stefan, Ljubljana, Slovenia, IJS-DP-8096, July 1999.
[6] NEA Nuclear Science Committee: "International Handbook of Evaluated Criticality Safety Benchmark Experiments, Nuclear Energy Agency", Organisation for Economic Co-operation and Development, NEA/NSC/DOC(95)03.
[7] Yo Makita and Akito Takahashi: "IAEA Benchmark Problem Based on the Time-of-Flight Experiment on Nickel Sphere at OKTAVIAN/Osaka University", IAEA, http://ripcnt01.iaea.or.at/nds/databases/fendl/FENDL.htm.
NEA-1553/48, included references:
[1] K. Sumita, A. Takahashi, T. Kasahara, et al: "Measurements of Neutron
Leakage Spectra from 16 cm Radius Nickel Sphere", OKTAVIAN Report A-84-04 (1984)
[2] A. Takahashi, J. Yamamoto, K. Oshima, et al: "Measurement of Double
Differential Neutron Emission Cross Sections for Fusion Reactor Candidate
Elements", Journal of Nuclear Science and Technology, Vol.21, No.8, 577-598
(1984)
[3] A. Takahashi, J. Yamamoto, H. Hashikura, et al: "Measurement and Analysis
of Neutron Leakage Spectrum from Nickel Sphere for 14 MeV Neutron Source",
Journal of Nuclear Science and Technology, Vol.23, No.6, 477-486 (1986)
[8] A. Milocco, The Quality Assessment of the OKTAVIAN Benchmark Experiments,
IJS-DP-10214, April 2009
[9] A. Milocco, A. Trkov, I. Kodeli: "The OKTAVIAN TOF Experiments in SINBAD:
Evaluation of the Experimental Uncertainties", Annals of Nuclear Energy 37
(2010) pp. 443-449

NEA-1553/50, bibliography:
SINBAD-OKTAVIAN/FE
==================
Background references:
[2] A. Takahashi, J. Yamamoto, H. Hashikura, et al: Experimental and Analytical Studies on Deep Penetration in Iron, OKTAVIAN Report C-84-09 (1984)
[3] K. Sumita, A. Takahashi, H. Hashikura, et al: Measurements of Neutron Leakage Spectra from 50.32 cm Radius Iron Sphere, UTNL-R 0159 (1983)
[4] A. Takahashi: Integral Neutronics Experiments at OKTAVIAN, OKTAVIAN Report B-83-01 (1983)
[6] M. Lanfranchi, J. F. Jaeger: Shielding Benchmarks. Some Considerations on Three Iron Shielding Benchmarks, EIR Report  TM-45-86-33, NEACRP Specialists' Meeting on Shielding Benchmarks, Paris (Oct. 1986)
[7] J. Hasnip, V. Herrnberger, Analysis of Fe-Shielding Benchmark Experiments by the JEF-1/EFF-Library, NEACRP Specialists' Meeting on Shielding Benchmarks, Paris (Oct. 1986)
[9] Fujio Maekawa, FNS/JAERI, Private communication (1998)
[10] A. Trkov, Comments on the Oktavian Iron Sphere Benchmark, Institute "Jozef Stefan", IJS-DP-7958 (Nov. 1998).
[11] Yo Makita and Akito Takahashi: IAEA Benchmark Problem Based on the Time-of-Flight Experiment on Iron Sphere at OKTAVIAN/Osaka University, IAEA, http://ripcnt01.iaea.or.at/nds/databases/fendl/FENDL.htm
NEA-1553/50, included references:
[1] K. Sumita, A. Takahashi, H. Hashikura, et al: Measurements of Neutron
Leakage Spectra from 50.32 cm Radius Iron Sphere, OKTAVIAN Report A-83-07 (1983)
[5] H. Hashikura, K. Haikawa, A. Takahashi, J. Yamamoto, K. Kanasugi, Y. Oka,
K. Sumita, S. An, Neutron Leakage Spectra From a Large Iron Sphere Pulsed with
14 MeV Neutrons, NEA Specialists' Meeting on Shielding Benchmarks, Saclay,
France (1984).
[8] A. Santamarina, I. Abidi, B. Gastaldi, Analysis of Fe Shield Benchmark
Experiments and Sensitivity Studies. JEF-EFF Data Base Improvement, JEF-EFF
Meeting, Paris (Dec. 1992)
[12] A. Milocco, The Quality Assessment of the OKTAVIAN Benchmark Experiments,
IJS-DP-10214, April 2009
[13] A. Milocco, A. Trkov, I. Kodeli: "The OKTAVIAN TOF Experiments in SINBAD:
Evaluation of the Experimental Uncertainties", Annals of Nuclear Energy 37
(2010) pp. 443-449

NEA-1553/52, bibliography:
SINBAD-OKTAVIAN/SI
==================
Background references:
[1] Ichihara C., et al.: Proc. Int. Conf. on Nucl. Data for Sci. and Technol., Mito, Japan, pp.319-322 (1988).
[2] Ichihara C., et al.: Proc. Second Specialists' Meeting on Nucl. Data for Fusion Reactors (1991), JAERI-M 91-062 (1991).
[3] Yamamoto J. et al.: "Gamma-Ray Emission Spectra from Spheres with 14 MeV Neutron Source", JAERI-M 89-026, 232 (1989).
[4] Yamamoto J. et al.: "Integral Experiment on Gamma-Ray Production at OKTAVIAN", JAERI-M 91-062, 118 (1991).
[5] Sumita K., et al.: Proc. 12th SOFT, Vol. 1 (1982)
[6] Yamamoto J. et al.: "Numerical Tables and Graphs of Leakage Neutron Spectra from Slabs of Typical Shielding Material with D-T Neutron Source", OKTAVIAN-Report A-8305, Dept. of Nuclear Eng., Osaka University (1983).
[7] Takahashi A., et el.: OKTAVIAN Report, C-83-02 (1983).
NEA-1553/52, included references:
[8] Sub Working Group of Fusion Reactor Physics Subcommittee: Collection of
Experimental Data for Fusion Neutronics Benchmark, JAERI-M-94-014, Feb. 1994.
[9] Fujio Maekawa, Masayuki Wada, Chihiro Ichihara, Yo Makita, Akito Takahashi,
Yukio Oyama: Compilation of Benchmark Results for Fusion Related Nuclear Data,
JAERI-Data/Code 98-024, Nov. 1998.
[10] A. Milocco, Quality Assessment of the OKTAVIAN Benchmark Experiments,
IJS-DP-10214, April 2009
[11] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D-T Neutron
Source in Ti-T Targets, IJS-DP-9988, July 2008
[12] A. Milocco, A. Trkov, I. Kodeli: "The OKTAVIAN TOF Experiments in SINBAD:
Evaluation of the Experimental Uncertainties", Annals of Nuclear Energy 37
(2010) pp. 443-449

NEA-1553/53, bibliography:
SINBAD-OKTAVIAN-W
=================
[1] Ichihara C., et al.: Proc. Int. Conf. on Nucl. Data for Sci. and Technol., Mito, Japan, pp.319-322 (1988).
[2] Ichihara C., et al.: Proc. Second Specialists' Meeting on Nucl. Data for Fusion Reactors (1991), JAERI-M 91-062 (1991).
[3] Yamamoto J. et al.: "Gamma-Ray Emission Spectra from Spheres with 14 MeV Neutron Source", JAERI-M 89-026, 232 (1989).
[4] Yamamoto J. et al.: "Integral Experiment on Gamma-Ray Production at OKTAVIAN", JAERI-M 91-062, 118 (1991).
[5] Sumita K., et al.: Proc. 12th SOFT, Vol. 1 (1982)
[6] Yamamoto J. et al.: "Numerical Tables and Graphs of Leakage Neutron Spectra from Slabs of Typical Shielding Material with D-T Neutron Source", OKTAVIAN-Report A-8305, Dept. of Nuclear Eng., Osaka University (1983).
[7] Takahashi A., et el.: OKTAVIAN Report, C-83-02 (1983).
[10] I. Kodeli, Recent Progress in the SINBAD Project, EFFDOC-866, EFF Meeting, Issy-les-Moulinaux (April 2003)
NEA-1553/53, included references:
[8] Sub Working Group of Fusion Reactor Physics Subcommittee: Collection of
Experimental Data for Fusion Neutronics Benchmark, JAERI-M-94-014, Feb. 1994.
[9] F. Maekawa, M. Wada, C. Ichihara, Y. Makita, A. Takahashi, Y. Oyama:
Compilation of Benchmark Results for Fusion Related Nuclear Data,
JAERI-Data/Code 98-024, Nov. 1998.
[11] A. Milocco, Quality Assessment of the OKTAVIAN Benchmark Experiments,
IJS-DP-10214, April 2009
[12] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D-T Neutron
Source in Ti-T Targets, IJS-DP-9988, July 2008
[13] A. Milocco, A. Trkov, I. Kodeli: "The OKTAVIAN TOF Experiments in SINBAD:
Evaluation of the Experimental Uncertainties", Annals of Nuclear Energy 37
(2010) pp. 443-449

NEA-1553/54, bibliography:
SINBAD-FNG-BLKT
===============
Background references:

[1] M. Martone, M. Angelone, M. Pillon, "The 14-MeV Frascati Neutron Generator
    (FNG)", ENEA Report RT/ERG/FUS/93/65
[2] M. Angelone, M. Pillon, P. Batistoni, M. Martini, M. Martone, V. Rado,
    " Absolute experimental and numerical calibration of the 14 MeV neutron
    source at the Frascati Neutron Generator,  Rev.Sci.Instrum. 67(6)(1996)2189
[3] P. Batistoni, M. Angelone, W. Daenner, U. Fischer, L. Petrizzi, M. Pillon,
    A. santamarina, K. Seidel, "Neutronics Shield Experiment for ITER at the
    Frascati Neutron Generator FNG", 17th Symposium on Fusion
    Technology, Lisboa, Portugal, September 16-20, 1996.
[4] J. H. Baard, W. L. Zijp, H. Nolthenius, "Nuclear Data Guide for Reactor
    Neutron Metrology",  Kluwer  Academic Publishers for the Commission of the
    European Community, (1989).
[5] M. Angelone, C. Arpesella, M. Martone, M. Pillon, Activation
    measurements for the E.C. bulck shield benchmark experiment, Proc. of
    the 4-th Int. Conf. on Application of Nuclear Techniques,
    Crete (Greece) June 1994, SPIE, Publ. vol. 2339 pp. 220-224
[6] M. Angelone, P. Batistoni. A. Esposito, M. Pillon, V. Rado, "Gamma
    and neutron dosimetry using CaF2:Tm thermoluminescent dosimeters for fusion
    reactor shielding experiments", Nucl. Sci. Eng. 126, p 176 June (1997).
NEA-1553/54, included references:
[7] L. Petrizzi, V. Rado, M. Angelone, P. Batistoni, M. Martone, M. Pillon,
    Analysis of nuclear heating experiments for the ITER shielding blanket,
    Proc. of  a Specialist Meeting on Measurement, Calculation and Evaluation of
    Photon Production  Data, Bologna, Italy Nov. 9-11, 1994,
    Edited by: C. Coceva,
    A. Mengoni, A. Ventura, NEA/NSC/DOC(95)1
[8] A. Milocco, The Quality Assessment of the FNG/TUD Benchmark Experiments,
    IJS-DP-10216, April 2009
[9] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron
    Source in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/55, bibliography:
SINBAD-FNG-DOSE-RATE
====================
Background references:

[4] P. Batistoni, S. Rollet, Y. Chen, U. Fischer, L. Petrizzi, Y.
     Morimoto, ?Analysis of dose rate experiment : comparison between
     FENDL, EFF/EAF and JENDL nuclear data libraries?, SOFT 2002
[5] M. Martone, M. Angelone, M. Pillon, The 14 MeV Frascati Neutron
     Generator, Journal of Nuclear Materials 212-215 (1994) 1661-1664;
[7] Briesmeister, J. F. (Ed.), MCNP - A general Monte Carlo n-particle
     transport code, version 4C, Report LA12625, Los Alamos, Sept. 1999.
[8] S. Ganesan and P. K. McLaughlin, FENDL/E - evaluated nuclear data
     library of neutron interaction cross-sections and photon production
     cross-sections and photon-atom interaction cross-sections for fusion
     applications, version 1.0, Report IAEA-NDS-128, Vienna, May 1994.
[9] R. A. Forrest, J-Ch. Sublet, ?FISPACT-99: User manual?, Report UKAEA
     FUS 407, December 1998
[10] M. Herman, A. B. Pashchenko, Extension and improvement of the FENDL
     library for fusion applications (FENDL-2), Report INDC(NDS)-373, IAEA
     Vienna, 1997.
[11] N. P. Kocherov, P. K. McLaughlin, The International Reactor Dosimetry
     File (IRDF-90), Report IAEA-NDS-141, Rev. 2, Oct. 1993.
NEA-1553/55, included references:
[1] P. Batistoni, M. Angelone, L. Petrizzi, M. Pillon, H. Freiesleben,
     D. Richter, K. Seidel, S. Unholzer, Y. Chen, U. Fischer, Experimental
     Validation of Shut-Down Dose Rates, Final Report, June 2001
[2] P. Batistoni, M. Angelone, L. Petrizzi, M. Pillon, ?Benchmark
     Experiment for the validation of shut down activation and dose
     calculation in a fusion device?, Journal of Nuclear Science and
     Technology, Sup. 2, p. 974-977 (August 2001), ND2001.
[3] P. Batistoni, L. Petrizzi, Task T426 - Neutronics Experiments,
     Experimental Validation of Shut Down Dose Rates, EFF-Doc-726, March 2000
[6] M. Angelone, P. Batistoni, M. Pillon, and V. Rado:
     Gamma and Neutron Dosimetry using CaF2:Tm Thermoluminescent
     Dosimeters for Fusion Reactor Shielding Experiments (EFF-Doc-614 (1997))
[12] A. Milocco, The Quality Assessment of the FNG/TUD Benchmark Experiments,
     IJS-DP-10216, April 2009
[13] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron
     Source in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/56, bibliography:
SINBAD-FNG-SIC
==============
Background references:
[5] M. Martone, M. Angelone, M. Pillon, The 14 MeV Frascati Neutron
    Generator, Journal of Nuclear Materials 212-215 (1994) 1661-1664;
[6] Briesmeister, J. F. (Ed.), MCNP - A general Monte Carlo n-particle
    transport code, version 4C, Report LA12625, Los Alamos, September 1999.
[7] M. Herman, A. B. Pashchenko, Extension and improvement of the FENDL
    library for fusion applications (FENDL-2), Report INDC(NDS)-373, IAEA
    Vienna, 1997.
[8] N. P. Kocherov, P. K. McLaughlin, The International Reactor Dosimetry
    File (IRDF-90), Report IAEA-NDS-141, Rev. 2, Oct. 1993.
[9] J.A. Gibson, The relative tissue-kerma sensitivity of thermoluminescent
    materials to neutrons, Report EUR 10105 EN (1985)
NEA-1553/56, included references:
[1] P. Batistoni, M. Angelone, L. Petrizzi, M. Pillon, Measurements and
    Analysis of Reaction Rates and of Nuclear Heating in SiC, Dec. 2001
[2] M. Angelone, P. Batistoni, I. Kodeli, L. Petrizzi, M. Pillon,
    ?Benchmark analysis of neutronics performances of a SiC block
    irradiated with 14 Mev neutrons?, Fus. Eng. Design 63-64 (2002) 475
[3] Y. Chen, U. Fischer, I. Kodeli, R. L. Perel, M. Angelone, P. Batistoni,
    L. Petrizzi, K. Seidel, S. Unholzer, ?Sensitivity and uncertainty
    analyses of 14 Mev neutron benchmark experiment on Silicon Carbide?,
    22nd Symposium on Fusion Technology, Helsinki, Finland, 9-13 Sept 2002.
[4] I. Kodeli, Deterministic Transport, Sensitivity and Uncertainty
    Analysis of SiC Benchmark Experiment Using EFF-3 and FENDL-2 Evaluations,
    EFFDOC-818, Dec. 2001.

NEA-1553/57, bibliography:
SINBAD-FNG-SS
=============
Background references:
[1] M. Martone, M. Angelone, M. Pillon, "The 14-MeV Frascati Neutron Generator
   (FNG)", ENEA Report RT/ERG/FUS/93/65
[2] J. H. Baard, W. L. Zijp, H. Nolthenius, "Nuclear Data Guide for Reactor
    Neutron Metrology", Kluwer Academic Publishers for the Commission of
    the European Community (1989).
NEA-1553/57, included references:
[3] A. Milocco, The Quality Assessment of the FNG/TUD Benchmark Experiments,
    IJS-DP-10216, April 2009
[4] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron Source
    in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/58, bibliography:
SINBAD-FNG-STREAMING
====================
Background references:
[1] M. Martone, M. Angelone, M. Pillon, The 14 MeV Frascati Neutron
    Generator, Journal of Nuclear Materials 212-215 (1994) 1661-1664;
[2] J.H. Hubble, Photon mass attenuation and energy-absorbtion
    coefficients from 1 keV up to 20 MeV, Int. J. Appl. Rad. Isot. 33
    (1982) 1269
[3] P. Batistoni, M. Angelone, L. Petrizzi, M. Pillon, "Neutron streaming
    experiment for ITER bulk shield at the Frascati 14-MeV neutron
    generator", Proceedings of the 20th Symposium on Fusion Technology,
    Marseille, France, 7-11 September 1998, Edited by B. Beaumont, P.
    Libeyre, B. de Gentile, G. Tonon, CEA Cadarache 1998, Vol.2, pag. 1417
[6] K. Seidel, M. Angelone, P. Batistoni et al., "Investigation of
    neutron and photon flux spectra in a streaming mock-up for ITER",
    Fusion Engineering and Design 51-52, (2000) 855-861
[10] J. F. Briesmeister (Ed.), MCNP - A General Monte Carlo N-Particle
     Transport Code, Version 4B, Report, Los Alamos National Laboratory,
     LA-12625-M, March 1997.
[11] S. Ganesan and P. K. McLaughlin, FENDL/E - evaluated nuclear data
     library of neutron interaction cross-sections and photon production
     cross-sections and photon-atom interaction cross-sections for fusion
     applications, version 1.0, Report IAEA-NDS-128, Vienna, May 1994.
[12] M. Herman, A. B. Pashchenko, Extension and improvement of the FENDL
     library for fusion applications (FENDL-2), Report INDC(NDS)-373, IAEA
     Vienna, 1997.
[13] A. J. Koning, H. Gruppelaar, A. Hogenbirk, Fusion Eng. Des. 37 (1997)
     211-216.
[14] N. P. Kocherov, P. K. McLaughlin, The International Reactor Dosimetry
     File (IRDF-90), Report IAEA-NDS-141, Rev. 2, Oct. 1993.
NEA-1553/58, included references:
[4] M. Angelone, P. Batistoni, L. Petrizzi, M. Pillon, "Neutron streaming
     Experiment at FNG: results and analysis", Fusion Engineering and
     Design 51-52, (2000) 653-661
[5] L. Petrizzi, P. Batistoni, I. Kodeli, "Sensitivity and uncertainty
     analysis performed on a 14-MeV neutron streaming experiment", Fusion
     Engineering and Design 51-52, (2000) 843-848
[8] P. Batistoni et al, ITER Task T.362: Neutron Streaming Experiment - Final
     Report, EFF-Doc-639, July 1998
[9] P. Batistoni, L. Petrizzi, Analysis of the Neutron Streaming Experiment
     using FENDL-1.0/2.0 and EFF-3.0/3.1 nuclear data libraries, EFF-DOC-673
[15] I. Kodeli, Report on the 1999 Activity on ND-1.2.1 Subtask: Processed
     Multigroup Covariance Files with extended options for EFF-3, EFFDOC-698
[16] A. Milocco, The Quality Assessment of the FNG/TUD Benchmark Experiments,
     IJS-DP-10216, April 2009
[17] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron
     Source in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/59, bibliography:
SINBAD-FNG-W
============
Backgroung references:
[2] P.Batistoni, M. Angelone, L. Petrizzi and M.Pillon, Neutronics
     Benchmark Experiment on Tungsten, presented at ICFRM-11 (2003),
     to be published in Journal of Nuclear Materials
[3] M. Martone, M. Angelone, M. Pillon, The 14 MeV Frascati Neutron
     Generator, Journal of Nuclear Materials 212-215 (1994) 1661-1664;
[4] J.H. Hubble, Photon mass attenuation and energy-absorption
     coefficients from 1 keV up to 20 MeV, Int. J. Appl. Rad. Isot.
     33 (1982) 1269
[5] Briesmeister, J. F. (Ed.), MCNP - A general Monte Carlo
     n-particle transport code, version 4C, Report LA12625, Los
     Alamos, September 1999.
[6] M. Herman, A. B. Pashchenko, Extension and improvement of the
     FENDL library for fusion applications (FENDL-2), Report
     INDC(NDS)-373, IAEA, Vienna, 1997.
[7] N. P. Kocherov, P. K. McLaughlin, The International Reactor
     Dosimetry File (IRDF-90), Report IAEA-NDS-141, Rev. 2, Oct. 1993.
NEA-1553/59, included references:
[1] P. Batistoni, M. Angelone, L. Petrizzi, M. Pillon, Measurements
     and Analysis of Neutron Reaction Rates and of Gamma Heating in
     Tungsten, MA-NE-R-003, ENEA, Dec. 2002
[8] I. Kodeli, Analysis of Benchmark Experiment on Tungsten Using
     DORT, TWODANT and SUSD3D Deterministic Analysis Tools,
     EFF Meeting, Issy-les-Moulinaux (Jan. 2003)
[9] I. Kodeli, Analysis of FNG Benchmark Experiment on Tungsten Using DORT,
     TWODANT and SUSD3D Deterministic Codes, EFFDOC-867 (April 2003)
[10] I. Kodeli, Tungsten Benchmark Experiments: Re-analysis Using JENDL-3.3,
     EFFDOC-885 (Nov. 2003)
[11] I. Kodeli, Cross-Section Sensitivity Analysis of 14 MeV Neutron
     Benchmark Experiment on Tungsten, Journal of Nuclear Materials,
     Vol. 329-333, Part.1, P. 717-720 (2004)
[12] A. Milocco, The Quality Assessment of the FNG/TUD Benchmark Experiments,
     IJS-DP-10216, April 2009
[13] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron
     Source in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/60, bibliography:
SINBAD-FNS-DUCT
===============
Background References:
[2] M. Matzke and K. Weise, "Neutron spectrum unfolding by the Monte
    Carlo method", Nucl. Inst. Meth., A234, 324 (1985).
[3] J.F. Briesmeister (Ed.), MCNP - a general Monte Carlo n-particle
    transport code, version 4C, LA-13709-M, Los Alamos National
    Laboratory (2000).
[4] A. B. Pashchenko, Summary Report of IAEA Consultants' Meeting on
    Selection of Basic Evaluations for the FENDL-2 Library,
    INDC(NDS)-356 (1996).
[5] K. Shibata, et al., "Japanese Evaluated Nuclear Data Library
    Version 3 Revision-3: JENDL-3.3," J. Nucl. Sci. Technol., 39,
    1125 (2002).
[6] F. Maekawa, C. Konno, et al., "Investigation of Prediction
    Capability of Nuclear Design Parameters for Gap Configuration in
    ITER through Analysis of the FNS Gap Streaming Experiment",
    J. Nucl. Sci. Technol., Supplement 1, 263 (2000).
[7] C. Konno, F. Maekawa, et al., "Experimental Investigation on
    Streaming due to a Gap between Blanket Modules in ITER", J. Nucl.
    Sci. Technol., Supplement 1, 540 (2000).
[8] C. Konno, F. Maekawa, et al., "Overview of Straight Duct Streaming
    Experiments for ITER", Fusion Eng. Des., 51-52, 797 (2000).
NEA-1553/60, included references:
[1] Y. Morimoto, K. Ochiai, T. Nishio, M. Wada, M. Yamauchi and
     T. Nishitani, "Dogleg Duct Streaming Experiment with 14 MeV Neutron
     Source", J. Nucl. Sci. Technol., supplement 4, p. 42-45 (March 2004).
[9] A. Milocco, The Quality Assessment of the FNS Benchmark Experiments,
     IJS-DP-10215, April 2009
[10] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron
     Source in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/61, bibliography:
SINBAD-FNS-OXYGEN
=================
Background references:
[1] Oyama Y., Maekawa H.: "Measurement and Analysis of an Angular Neutron
    Flux on a Beryllium Slab Irradiated with Deuteron-Tritium Neutrons,"
    Nucl. Sci. Eng., 97, 220-234 (1987).
[5] Oyama Y., Maekawa H.: "Spectral Measurement of Angular Neutron Flux on
    the Restricted Surface of Slab Assemblies by the Time-of-Flight Method,"
    Nucl. Instr. Methods, A245 173-181 (1986).
[6] Oyama Y., Yamaguchi S., Maekawa H.: "Measurements and Analyses of
    Angular Neutron Flux Spectra on Graphite and Lithium-Oxide Slabs
    Irradiated with 14.8 MeV Neutrons," J. Nucl. Sci. Technol., 25,
    419-428 (1988).
[9] International Atomic Energy Agency, Nuclear Data Section:
    Compilation for FENDL benchmarks,
    "http://ripcnt01.iaea.org/nds/databases/fendl/fen-bench.htm".
NEA-1553/61, included references:
[2] Oyama Y., Yamaguchi S., Maekawa H.: "Experimental Results of Angular
    Neutron Flux Spectra Leaking from Slabs of Fusion Reactor Candidate
    Materials (I)," JAERI-M 90-092 (1990).
[3] Oyama Y., Maekawa H.: "Measurements of Angle-Dependent Neutron
    Spectra from Lithium-Oxide Slab Assemblies by Time-of-Flight Method,"
    JAERI-M 83-195 (Nov. 1983).
[4] Oyama Y., Yamaguchi S., Maekawa H.: "Analysis of Time-of-Flight
    Experiment on Lithium-Oxide Assemblies by a Two-Dimensional
    Transport Code DOT3.5," JAERI-M 85-031 (March 1985).
[7] Oyama Y., Kosako K., Maekawa H.: "Measurements and Analyses of
    Angular Neutron Flux Spectra on Liquid Nitrogen, Liquid Oxygen and
    Iron Slabs," Proc. Int'l Conf. on Nuclear Data for Science and
    Technology, 13-17 May, Juelich (1991).
[8] Oyama Y., Kosako K., Maekawa H.: "Measurement and Calculations of
    Angular Neutron Flux Spectra Leaking from Slabs Bombarded with 14.8 MeV
    Neutrons," Nucl. Sci. Eng., 115, 24-37 (1993)
[10] F. Maekawa, M. Wada, C. Ichihara, Y. Makita, A. Takahashi, Y. Oyama:
     Compilation of Benchmark Results for Fusion Related Nuclear Data,
     JAERI-Data/Code 98-024, Nov. 1998.
[11] Y. Oyama: "Experimental Study of Angular Neutron Flux Spectra
     on a Slab Surface to Assess Nuclear Data and Calculational Methods
     for a Fusion Reactor Design", JAERI-M 88-101 (June 1988).
[12] A. Milocco, The Quality Assessment of the FNS Benchmark Experiments,
     IJS-DP-10215, April 2009
NEA-1553/62, included references:
[1] S. Yoshida, T. Nishitani, K. Ochiai, J. Kaneko, J. Hori, S. Sato,
    M. Yamauchi, R. Tanaka, M. Nakao, M. Wada, M. Wakisaka, I. Murata,
    C. Kutukake, S. Tanaka, T. Sawamura, and A. Takahashi:
    Measurement of Radiation Skyshine with D-T Neutron Source,
    Fus. Eng. Des. 69 (2003) 637-641
    (22nd Symposium on Fusion Technology (SOFT-22), Helsinki, 9-13. Sept. 2002).
[2] T. Nishitani, K. Ochiai, S. Yoshida, R. Tanaka, M. Wakisaka, M. Nakao,
    S. Sato, M. Yamauchi, J. Hori, M. Wada, A. Takahashi, J. Kaneko and
    T. Sawamura, D-T Neutron Skyshine Experiments and the MCNP Analysis,
    J. Nucl. Sci. & Technol., Sup. 4, p. 58-61 (March 2004)
[3] A. Milocco, The Quality Assessment of the FNS Benchmark Experiments,
    IJS-DP-10215, April 2009

NEA-1553/69, bibliography:
SINBAD-TUD-FNG-BS
=================
Background references:
[1] M. Martone, M. Angelone, M. Pillon, The 14 MeV Frascati Neutron Generator, Journal of Nuclear Materials 212-215 (1994) 1661-1664; M. Pillon, M. Angelone, A. V. Krasilnikov, 14 MeV Neutron Spectra Measurements with 4% Energy Resolution using Type IIa Diamond Detector,Nucl. Instr. Meth. in Phys. Res. B101 (1995) 473-485.
[2] H. Freiesleben, W. Hansen, D. Richter, K. Seidel, S. Unholzer, P.Batistoni, M. Pillon, M. Angelone, Investigation of Neutron and Gamma-ray Spectra in a Blanket Mock-up of the International Thermonuclear Experimental Reactor (ITER), Proc. of the 9th Intern.Symp. on Reactor Dosimetry, Prague, Czech Republic, 2-6 September
1996, Editors: H. Ait Abderrahim, P. D'hondt and B. Osmera, World Scientific, Singapore, 1998, p. 391-396.
[3] H. Freiesleben, W. Hansen, D. Richter, K. Seidel, S. Unholzer, U. Fischer, Y. Wu, M. Angelone, P. Batistoni, M. Pillon, Measurement and Analysis of Spectral Neutron and Photon Fluxes in an ITER Shield Mock-Up, Fusion Technology 1996, Proc. of the 19th. Symp. on Fusion Technology, Lisbon, Portugal, 16-20 September 1996, C. Varandas and F. Serra (editors), Elsevier Science B.V., Amsterdam, 1997, p. 1571-1574.
[6] U. Fischer, H. Freiesleben, W. Hansen, D. Richter, K. Seidel, S. Unholzer, Y. Wu, Test of evaluated data from libraries for fusion applications in an ITER shield mock-up experiment, International Conference on Nuclear Data for Science and Technology, Trieste, May 19-24, 1997, Conference Proceedings Vol. 59, p. 1215-1217, G. Reffo, A. Ventura and C. Grandi (Eds.), SIF, Bologna, 1997.
[7] M. Tichy, The DIFBAS Program - Description and User's Guide, Report PTB-7.2- 193-1, Braunschweig 1993.
[8] S. Guldbakke, H. Klein, A. Meister, J. Pulpan, U. Scheler, M. Tichy, S. Unholzer, Response Matrices of NE213 Scintillation Detectors for Neutrons, Reactor Dosimetry ASTM STP 1228, Ed. H. Farrar et al., American Society for Testing Materials, Philadelphia, 1995, p. 310-322.
[9] J. F. Briesmeister (Ed.), MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A, Report, Los Alamos National Laboratory, LA-12625-M, November 1993.
[10] S. Ganesan and P. K. McLaughlin, FENDL/E - evaluated nuclear data library of neutron interaction cross-sections and photon production cross-sections and photon-atom interaction cross-sections for fusion applications, version 1.0, Report IAEA-NDS-128, Vienna, May 1994.
[11] M. Herman, A. B. Pashchenko, Extension and improvement of the FENDL library for fusion applications (FENDL-2), Report INDC(NDS)-373, IAEA Vienna, 1997.
NEA-1553/69, included references:
[4] H. Freiesleben, W. Hansen, D. Richter, K. Seidel, S. Unholzer, U. Fischer,
Y. Wu, M. Angelone, P. Batistoni, M. Pillon, Measurement of Neutron and Gamma
Spectral Fluxes in the Shielding Assembly, Report TU Dresden, Institut fuer
Kern- und Teilchenphysik, TUD-IKTP/96-04, November 1996.
[5] H. Freiesleben, W. Hansen, D. Richter, K. Seidel, S. Unholzer, U. Fischer,
Y. Wu, M. Angelone, P. Batistoni, M. Pillon, Neutron and Photon Flux Spectra in
a Mock-up of the ITER Shielding System, Fusion Engineering and Design 42
(1998), Proc. of the Fourth Intern. Symp. on Fusion Nuclear Technology, Tokyo,
April 6-11, 1997, M.A. Abdou (Ed.), Elsevier Science B.V., Part C, p. 247-253.
[12] U. Fischer, Y. Wu, W. Hansen, D. Richter, K. Seidel, S. Unholzer,
Benchmark Analyses for the ITER Bulk Shield Experiment with EFF-3.0, -3.1 and
FENDL-1, -2 Nuclear Cross-Section Data, IAEA FENDL-2 Consultants' Meeting,
October 12-14, 1998, Vienna.
[13] I. Kodeli, Report on 1999 Activity on ND-1.2.1 (extracts), EFF/DOC-698,
EFF Meeting, Issy-les-Moulineaux NEA-DB (Nov. 1999)
[14] P. Batistoni, M. Angelone, U. Fischer, H. Freiesleben, W. Hansen, M.
Pillon, L. Petrizzi, D. Richter, K. Seidel, S. Unterholzer: Neutronics
Experiment on a Mock-up of the ITER Shielding Blanket at the Frascati Neutron
Generator, Fusion Engineering Design 47 (1999) 25-60
[15] A. Milocco, The Quality Assessment of the FNG/TUD Benchmark Experiments,
IJS-DP-10216, April 2009
[16] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron
Source in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/70, bibliography:
SINBAD-TUD-FNG-SIC
==============
Background references:
[1] M. Angelone, M. Pillon, P. Batistoni, M. Martini, M. Martone, V. Rado, "Absolute experimental and numerical calibration of the 14 MeV neutron source at the Frascati Neutron Generator", Rev. Sci. Instr. 67(1996)2189.
[2] M. Tichy, "The DIFBAS Program - Description and User's Guide", Report PTB-7.2- 193-1, Braunschweig 1993.
[3] S. Guldbakke, H. Klein, A. Meister, J. Pulpan, U. Scheler, M. Tichy, S. Unholzer, "Response Matrices of NE213 Scintillation Detectors for Neutrons", Reactor Dosimetry ASTM STP 1228, Ed. H. Farrar et al., American Society for Testing Materials, Philadelphia, 1995, p. 310.
[4] J. F. Briesmeister (Ed.), "MCNP - A general Monte Carlo n-particle transport code", version 4C, Report LA-13709, Los Alamos National Laboratory, 2000.
[5] H. Wienke, M. Herman, "FENDL/MG-2.0 and FENDL/MC-2.0 - The processed cross section libraries for neutron and photon transport calculations", Report IAEA-NDS-128, Vienna, 1998.
[6] S. Tagesen, H. Vonach, "Evaluation of neutron cross sections for fusion relevant materials", EFFDOC-785, NEA Data Bank, Nov. 2001.
NEA-1553/70, included references:
[7] H. Freiesleben, C. Negoita, K. Seidel, S. Unholzer, Y. Chen, U. Fischer
R. L. Perel, M. Angelone, P. Batistoni, M. Pillon:
Measurement and analysis of neutron and gamma-ray flux spectra in Tungsten
Report TUD-IKTP/02-02, Dresden, 2002, EFFDOC-822
[8] U. Fischer, R. Perel and Y. Chen
Monte Carlo Transport Sensitivity and Uncertainty Analyses for the TUD
Benchmark Experiment on SiC, EFFDOC-815
[9] K. Seidel, M. Angelone, P. Batistoni, Y. Chen, U. Fischer, H. Freiesleben
C. Negoita, R. L. Perel, M. Pillon, S. Unholzer
Measurement and Analysis of Neutron and Gamma-Ray Flux Spectra in Sic
Fus. Eng. Design 69 (2003) 379
[10] Y. Chen, U. Fischer, I. Kodeli, R. L. Perel, M. Angelone, P. Batistoni
L. Petrizzi, K. Seidel, S. Unholzer
Sensitivity and uncertainty analyses of 14 Mev neutron benchmark experiment
on Silicon Carbide, Fus. Eng. Design 69 (2003) 437-442
[11] A. Milocco, The Quality Assessment of the FNG/TUD Benchmark Experiments,
IJS-DP-10216, April 2009
[12] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron
Source in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/71, bibliography:
SINBAD-FNG-HCPB
===============
Background references:
[3] M. Martone, M. Angelone, M. Pillon, The 14 MeV Frascati Neutron Generator, Journal of Nuclear Materials 212-215 (1994) 1661-1664;
[4] P. Batistoni, Status of the Neutronics Experiment on a Mock-up of a Test Blanket Module (TBM), EFF-DOC-896 (2004)
[5] P. Batistoni, P. Carconi, M. Angelone, G. Zappa, Design of TBM Neutronics Experiment, Part 2: Design of the Measurements of Tritium Production and of Nuclear Heating in the Mock-up: Benchmarking of Experimental Techniques, Assessment of Uncertainties, FUS TEC MA-NE- R - 008 (2003).
NEA-1553/71, included references:
[1] P. Batistoni, R. Villari, TBM - HCPB Neutronics Experiments: Comparison and
Check Consistency among Results Obtained by the Different Teams, Implications
for ITER TBM Nuclear Design and Final Assessment, FUS-TEC-MA-NE-R-019, ENEA,
Dec. 2006
[2] P.Batistoni, P. Carconi, R. Villari, M. Angelone, M. Pillon, G. Zappa,
Measurements and Analysis of Tritium Production Rate (TPR) in Ceramic Breeder
and of Neutron Flux by Activation Rates in Beryllium in TBM Mock-up,
FUS-TEC-MA-NE-R-014, Dec. 2005
[6] P. Batistoni, Status of TBM Neutronics Experiment, EFF-DOC-938,  EFF/EAF
Meeting, NEA Data Bank, Paris, 28 December 2005
[7] S. Villari, P. Batistoni, M. Angelone, Status of the HCPB-TBM Benchmark
Experiment, TBM Neutronics Experiment Meeting, Frascati, 12 Sept. 2005
[8] D. Leichtle, U. Fischer, I. Kodeli, R. L. Perel, M. Angelone, P.
Batistoni, P. Carconi, M. Pillon, I. Schaefer, K. Seidel, R. Villari, G. Zappa,
"Sensitivity and Uncertainty Analyses of the Tritium Production in the HCPB
Breeder Blanket Mock-up Experiment", Fusion Engineering and Design, 82 (15),
p.2406-2412 (2007)
[9] P. Batistoni, M. Angelone, L. Bettinali, P. Carconi, U. Fischer, I.
Kodeli, D. Leichtle, K. Ochiai, R. Perel, M. Pillon, I. Schaefer, K. Seidel, Y.
Verzilov, R. Villari, G. Zappa, "Neutronics Experiment on a HCPB Breeder
Blanket Mock-up", Fusion Engineering and Design, 82 (15)}, p.2095-2104, (2007)
[10] I. Kodeli, 2D and 3D Deterministic Transport, Sensitivity and Uncertainty
Analysis of HCPB Tritium Breeder Module Mock-up Benchmark, IJS-DP-9312, January
2006
[11] I. Kodeli, Deterministic 3D Transport, Sensitivity and Uncertainty
Analysis of TPR and Reaction Rate Measurements in HCPB Breeder Blanket Mock-up
Benchmark, EFF Meeting, OECD/NEA, 22 May 2006, EFF-DOC-981
[12] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D-T Neutron
Source in Ti-T Targets, IJS-DP-9988, July 2008
[13] I. Kodeli, Deterministic 3D Transport, Sensitivity and  Uncertainty
Analysis of TPR and Reaction Rate Measurements in HCPB Breeder Blanket Mock-up
Benchmark, Nuclear Energy for New Europe 2006 International Conference
(Portoroz, Slovenia, September 18-21, 2006)

NEA-1553/72, bibliography:
SINBAD-FNS-C-CYLIND
===================
Background references:
[1] Maekawa H., et al.: "Fusion Blanket Benchmark Experiments on a 60 cm-Thick Lithium Oxide Cylindrical Assembly," JAERI-M 86-182 (1986).
[3] Seki M., et al.: J. Nucl. Sci. Technol., 16, 8238 (1979).
[4] Maekawa H., et al.: "Neutron Yield Monitor for the Fusion Neutronics Source (FNS) For 80 deg. Beam Line ---," JAERI-M 83-219 (1983).
[5] Oyama Y. and Maekawa H.: Nucl. Instr. Meth., A245, 173 (1986).
[6] Oyama Y., et al.: "Development of a Spherical NE213 Spectrometer with 14 mm Diameter," JAERI-M 84-124 (in Japanese, 1984); Nucl. Instr. Meth., A256, 333 (1987).
[7] Seki Y., et al.: J. Nucl. Sci. Technol., 20, 686 (1983).
[10] Maekawa H., et al.: "Integral Experiment on Graphite Cylindrical Assembly", International Atomic Energy Agency, Nuclear Data Section: Compilation for FENDL benchmarks, available from: http://ripcnt01.iaea.org/nds/databases/fendl/fen-bench.htm
NEA-1553/72, included references:
[2] Maekawa H., et al.: "Benchmark Experiments on a 60 cm-Thick Graphite
Cylindrical Assembly," JAERI-M 88-034 (1988)
[8] F. Maekawa, J. Yamamoto, C. Ichihara, K. Ueki, Y. Ikeda: "Sub Working Group
of Fusion Reactor Physics Subcommittee: Collection of Experimental Data for
Fusion Neutronics Benchmark", JAERI-M-94-014, Feb. 1994 (Ch. 1.6, p.217)
[9] Fujio Maekawa, Masayuki Wada, Chihiro Ichihara, Yo Makita, Akito Takahashi,
Yukio Oyama: "Compilation of Benchmark Results for Fusion Related Nuclear
Data", JAERI-Data/Code 98-024, Nov. 1998.
[11] A. Milocco, The Quality Assessment of the FNS Benchmark Experiments,
IJS-DP-10215, April 2009

NEA-1553/73, bibliography:
SINBAD-FNS/V
============
Background references:
[3] F. Maekawa, C. Konno, K. Kosako, Y. Oyama, Y. Ikeda and H. Maekawa: Bulk Shielding Experiments on Large SS316 Assemblies Bombarded by D-T Neutrons, Volume II: Analysis, JAERI-Research 94-044 (1994).
[4] Y. Oyama: Experimental Study of Angular Neutron Flux Spectra on a Slab Surface to Assess Nuclear Data and Calculational Methods for a Fusion Reactor Design, JAERI-M 88-101 (1988).
NEA-1553/73, included references:
[1] F. Maekawa et al.:
Data Collection of Fusion Neutronics Benchmark Experiment conducted at FNS/JAERI
JAERI-Data/Code 98-021 (August 1998)
[2] F. Maekawa et al.:
Benchmark Experiment on Vanadium with D-T Neutrons and Validation of Evaluated
Nuclear Data Libraries by Analysis of the Experiment
Journal of Nuclear Science and Technology, Vol.36, No.3, p.242-249 (March 1999)
[5] A. Milocco, The Quality Assessment of the FNS Benchmark Experiments,
JS-DP-10215, April 2009
[6] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron Source
in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/74, bibliography:
SINBAD-FNS-W
============
Background references:
[3] F. Maekawa, C. Konno, K. Kosako, Y. Oyama, Y. Ikeda and H. Maekawa: Bulk Shielding Experiments on Large SS316 Assemblies Bombarded by D-T Neutrons, Volume II: Analysis, JAERI-Research 94-044 (1994).
[4] Y. Oyama: Experimental Study of Angular Neutron Flux Spectra on a Slab Surface to Assess Nuclear Data and Calculational Methods for a Fusion Reactor Design, JAERI-M 88-101 (1988).
[5] I. Kodeli, Recent Progress in the SINBAD Project, EFFDOC-866, EFF Meeting, Issy-les-Moulinaux (April 2003)
NEA-1553/74, included references:
[1] F. Maekawa, C. Konno, Y. Kasugai, Y. Oyama, Y. Ikeda:
Data Collection of Fusion Neutronics Benchmark Experiment Conducted at
FNS/JAERI, JAERI-Data/Code 98-021 (1998)
[2] F. Maekawa, M. Wada, C. Ichihara, Y. Makita, A.Takahashi, Y. Oyama:
Compilation of Benchmark Results for Fusion Related Nuclear Data,
JAERI-Data/Code 98-024 (1998)
[6] A. Milocco, The Quality Assessment of the FNS Benchmark Experiments,
IJS-DP-10215, April 2009
[7] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D?T Neutron Source
in Ti-T Targets, IJS-DP-9988, July 2008

NEA-1553/75, bibliography:
SINBAD-IPPE-FE
==============
[2] B.V. Devkin, M.G. Kobozev, S.P. Simakov, V.V. Sinitca, V.A. Talalaev, U. Fischer, U. von Mollendorff, E. Wiegner:
"Neutron Leakage Spectra from Iron Spheres". In: Fusion Technology 1994, ed. by K. Herschbach et al., vol. 2, p. 1357, Elsevier Science, Amsterdam, 1995.
[3] S.P. Simakov, B.V. Devkin, M.G. Kobozev, A.A. Lychagin, V.A Talalaev, A.A. Androsenko:
"14 MeV Facility and Research in IPPE", Report INDC(CCP)-351, IAEA, Vienna, 1993; Voprocy Atomnoy Nauki i Tehniki, Seriya Yadernye Konstanty, Obninsk, 1997, no. 3-4, p. 93.
[4] B.V. Devkin, M. G. Kobozev, S.P. Simakov, U. Fischer, F. Kappler, U. von Mollendorff:
"Evaluation of Corrections for Spherical-Shell Neutron Transmission Experiments by the Monte-Carlo Technique", Report FZKA 5862, Karlsruhe, 1996; Voprocy Atomnoy Nauki i Tehniki, Seriya Yadernye Konstanty, Obninsk, 1997, no. 1-2, p. 38.
NEA-1553/75, included references:
[1] S.P. Simakov, B.V. Devkin, M.G. Kobozev, V.A. Talalaev, U. Fischer, U. von
Moellendorff: "Validation of evaluated data libraries against an iron shell
transmission experiment and against the Fe(n,xn) reaction cross section with 14
MeV neutron source", Report EFF-DOC-747, EFF/EAF Fusion Nuclear Data and
Neutronics Meeting, 5-6 December 2000, NEA Data Bank, Paris
[5] A. Milocco, "Quality Assessment of the IPPE Benchmark Experiments",
IJS-DP-10217, April 2009

NEA-1553/76, bibliography:
SINBAD-IPPE-V
=============
Background references:
[3] U. von Moellendorf et al., "A 14-MeV neutron transmission experiment on vanadium", 19th Symposium on Fusion Technology, Lisbon, 16-20 Sept. 1996.
NEA-1553/76, included references:
[1] S. P. Simakov et al.:
Benchmarking of evaluated nuclear data for
vanadium by a 14 MeV spherical shell transmission experiment
International data Committee, Report INDC(CCP)-417. October 1998
[2] S. P. Simakov et al.:
Benchmarking of evaluated nuclear data for
vanadium by a 14 MeV spherical shell transmission experiment
Forschungzentrum Karlsruhe, Report FZKA 6096, 1998
[4] A. Milocco, "Quality Assessment of the IPPE Benchmark Experiments",
IJS-DP-10217, April 2009

NEA-1553/77, bibliography:
SINBAD-OKTAVIAN/AL
==================
Background references:
[1] Ichihara C., et al.: Proc. Int. Conf. on Nucl. Data for Sci. and Technol., Mito, Japan, pp.319-322 (1988).
[2] Ichihara C., et al.: Proc. Second Specialists' Meeting on Nucl. Data for Fusion Reactors (1991), JAERI-M 91-062 (1991).
[3] Sumita K., et al.: Proc. 12th SOFT, Vol. 1 (1982)
[4] Takahashi A., et el.: OKTAVIAN Report, C-83-02 (1983).
[6] International Atomic Energy Agency, Nuclear Data Section: Compilation for FENDL benchmarks, http://ripcnt01.iaea.org/nds/databases/fendl/fen-bench.htm
NEA-1553/77, included references:
[5] Sub Working Group of Fusion Reactor Physics Subcommittee: Collection of
Experimental Data for Fusion Neutronics Benchmark, JAERI-M-94-014, Feb. 1994.
[7] Fujio Maekawa, Masayuki Wada, Chihiro Ichihara, Yo Makita, Akito Takahashi,
Yukio Oyama: Compilation of Benchmark Results for Fusion Related Nuclear Data,
JAERI-Data/Code 98-024, Nov. 1998.
[8] Sumita K., et el.: OKTAVIAN Report, C-83-01 (1983).
[9] A. Milocco, Quality Assessment of the OKTAVIAN Benchmark Experiments,
IJS-DP-10214, April 2009
[10] A. Milocco, A. Trkov, MCNPX/MCNP5 Routine for Simulating D-T Neutron Source
in Ti-T Targets, IJS-DP-9988, July 2008
[11] A. Milocco, A. Trkov, I. Kodeli: "The OKTAVIAN TOF Experiments in SINBAD:
Evaluation of the Experimental Uncertainties", Annals of Nuclear Energy 37
(2010) pp. 443-449

NEA-1553/78, bibliography:
[5] Takahashi A., et el.: OKTAVIAN Report, C-83-02 (1983).
[6] I. Kodeli, Recent Progress in the SINBAD Project, EFFDOC-866, EFF Meeting, Issy-les-Moulinaux (April 2003).
NEA-1553/78, included references:
[1] Sub Working Group of Fusion Reactor Physics Subcommittee: Collection of
Experimental Data for Fusion Neutronics Benchmark, JAERI-M-94-014, Feb. 1994.
[2] F. Maekawa, M. Wada, C. Ichihara, Y. Makita, A. Takahashi, Y. Oyama:
Compilation of Benchmark Results for Fusion Related Nuclear Data,
JAERI-Data/Code 98-024, Nov. 1998.
[3] A. Takahashi, J. Yamamoto, K. Oshima, et al: "Measurement of Double
Differential Neutron Emission Cross Sections for Fusion Reactor Candidate
Elements", Journal of Nuclear Science and Technology, Vol.21, No.8, 577-598
(1984).
[4] A. Milocco, Quality Assessment of the OKTAVIAN Benchmark Experiments,
IJS-DP-10214, April 2009.
top ]
12. PROGRAMMING LANGUAGE(S) USED

No item found

top ]
15. NAME AND ESTABLISHMENT OF AUTHORS
NEA-1553/26
SINBAD-ILL-FE
=============
Author/Organizer
----------------
Experiment and Analysis:
Richard Harold Johnson, Nuclear Engineering Program, University of Illinois at Urbana-Champaign, Urbana, IL, USA
  
Compiler of data for SINBAD:
Jennifer Parsons, Radiation Shielding Information Center, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6362, USA
  
Reviewer of compiled data:
Hamilton Hunter, Radiation Shielding Information Center, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6362, USA

NEA-1553/27
SINBAD-SB5-FUS
==============
Author/Organizer:
----------------
Authors:
G. T. Chapman, G. L. Morgan, J. W. McConnell, R. T. Santoro, J. M. Barnes, R. G. Alsmiller, Jr., and E. M. Oblow. Oak Ridge National Laboratory, USA
  
Compiler:
C. O. Slater, Oak Ridge National Laboratory, Oak Ridge, Tennessee USA 37831-6363.
  
Reviewer of compiled data:
Hamilton Hunter, Radiation Shielding Information Center, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6362, USA

NEA-1553/40
SINBAD-MEPHI
============
Author/Organizer
----------------
Experiment and Analysis:
Romodanov V.L.* - science lieder and physics method creation
Andreev M.I.* - heating rates measurements
Afanasiev V.V.* - reaction rates measurements
Belevitin A.G.* - reaction rates measurements
Sacharov V.K.* - MCNP-4c2 calculation models
Trykov L.A.** - neutron and photons spectra measurements
  
*Moscow Engineering-Physics Institute, State University Moscow, 115409, Kashirskoe Shosse 31, Russia
**State Research Center 'Institute of Nuclear Power Engineering',
Obninsk, Kaluga region, Bondarenko Square 2, Russia
  
Compiler of data for Sinbad:
Romodanov V.L.
Sacharov V.K.
Moscow Engineering-Physics Institute (State University),
Moscow, 115409, Kashirskoe Shosse 31, Russia
  
Reviewer of compiled data:
I. Kodeli
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
  
Acknowledgement
---------------
The experiment and the corresponding analysis was performed in the framework of ISTC project No. 180.

NEA-1553/41
SINBAD-KANT
===========
Author/Organizer:
----------------
Experiment and analysis:
U. von Moellendorff, H. Giese, F. Kappler, W. Eyrich, H. Fries,
Forschungszentrum Karlsruhe, Karlsruhe, Germany;
  
K. Hayashi, T. Tsukiyama, H. Ebi, R. Tayama,
Hitachi Engrg. Co., Hitachi-shi, Japan
  
Compiler of data for Sinbad:
U. Fischer
Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit,
Postfach 3640, D-76021 Karlsruhe, Germany;
  
Reviewer of compiled data:
E. Sartori, I. Kodeli
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France

NEA-1553/43
SINBAD-LI-BLANKET
=================
Author/Organizer:
Experiment and analysis:
G.Gyorgen, R.Herzing, L.Kuijpers, W.Pohorecki
  
Compiler of data for SINBAD:
Wladyslaw Pohorecki, Faculty of Physics and Applied Computer Sciences, AGH University of Science and Techniques, Krakow, Poland
TN +48 12 6172954, fax +40 12 6340010
e-mail: poho@a

Review of compiled data: to be carried out

NEA-1553/45
SINBAD-FNS-OXYGEN
=================
Author/Organizer:
----------------
Department of Reactor Engineering
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken
Japan
  
The compilation and review need to be carried out. Contributions are highly appreciated.

NEA-1553/46
SINBAD-TUD-FE
=============
Author/Organizer:
----------------
Experiment and analysis:
H. Freiesleben, W. Hansen, D. Richter, K. Seidel, S. Unholzer
Technische Universitaet Dresden
Institut fuer Kern- und Teilchenphysik
D-01062 Dresden
Germany
  
U. Fischer, Y. Wu
Forschungszentrum Karlsruhe
Institut fuer Neutronenphysik und Reaktortechnik
Postfach 3640
D-76021 Karlsruhe
Germany
  
Compiler of data for Sinbad:
K. Seidel
Technische Universitaet Dresden
Institut fuer Kern- und Teilchenphysik
D-01062 Dresden
Germany
  
Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia
  
Reviewer of compiled data:
A. Avery,
Performance and Safety Services Department,
AEA Technology
WINFRITH, Dorchester
Dorset DT2 8DH
UK

NEA-1553/47
SINBAD-TUD-FNG-W
============
Authors/Organizer:
------------------
Experiment and analysis:
K. Seidel, H. Freiesleben, C. Negoita, S. Unholzer
Technische Universitaet Dresden
Institut fuer Kern- und Teilchenphysik
D-01062 Dresden
Germany
  
U. Fischer, D. Leichtle
Forschungszentrum Karlsruhe
Institut fuer Reaktorsicherheit
D-76021 Karlsruhe
Germany
  
M. Angelone, P. Batistoni, M. Pillon
Associazione ENEA-EURATOM
Settore Fusione - Divisione Neutronica
Via E. Fermi 27
I-00044 Frascati (Rome)
Italy   
  
Compiler of data for Sinbad:
K. Seidel
Technische Universitaet Dresden
Institut fuer Kern- und Teilchenphysik
D-01062 Dresden
Germany
  
Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia
  
Reviewer of compiled data:
I. Kodeli
OECD/NEA
12 bd des Iles
92130 Issy les Moulineaux
France

NEA-1553/48
SINBAD-OKTAVIAN/NI
==================
Author/Organizer
----------------
Experiment and analysis:
Akito Takahashi,J. Yamamoto, K. Sumita
Faculty of Nuclear Engineering
Department of Engineering
Osaka University
2-1 Yamadaoka, Suita-shi Osaka-fu, 565 Japan
  
T. Kasahara, H. Hashikura, M. Akiyama, Y. Oka, S. Kondo
Nuclear Engineering Research Laboratory, Faculty of Engineering
University of Tokyo
Tokai-mura, Ibaraki 319-11, JAPAN
  
Compiler of data for Sinbad:
I. Kodeli
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
   
Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia
  
Reviewer of compiled data:
A. Trkov
Institute Jozef Stefan, Jamova 39, 1000 Ljubljana, Slovenia

NEA-1553/50
SINBAD-OKTAVIAN/FE
==================
Author/Organizer
----------------
Experiment and analysis:
H. Hashikura, K. Haikawa, A. Takahashi *, J. Yamamoto *,
K. Kanasugi, Y. Oka, K. Sumita *, S. An,
Nuclear Engineering Research Laboratory, Faculty of Engineering
University of Tokyo
Tokai-mura, Ibaraki 319-11, Japan
  
* Department of Nuclear Engineering
Osaka University
Yamada-oka, Suita, Osaka 564, Japan
  
Compiler of data for Sinbad:
I. Kodeli
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
  
Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia
  
Reviewer of compiled data:
A. Trkov
Institute Jozef Stefan, Jamova 39, 1000 Ljubljana, Slovenia

NEA-1553/52
SINBAD-OKTAVIAN/SI
==================
AUTHOR/ORGANISER
----------------
Experiment and analysis:
Chihiro Ichihara, Katsuhei Kobayashi:
Research Reactor Institute, Kyoto University
Noda, Sennan-gun, Osaka 590-04, Japan
  
Shu A. Hayashi:
Institute for Atomic Energy, Rikkyo University
2-5-1 Nagasaka, Yokosukas Kanagawa 240-01, Japan
  
Itsuro Kimura:
Department of Nuclear engineering, Faculty of Engineering,
Kyoto University
Yoshida-honmachi, Sakyo-ku, Kyoto 606, Japan
  
Junji Yamamoto, Akito Takahashi, T. Kanaoka, I. Murata,
K. Sumita:
Department of Nuclear Engineering, Faculty of Engineering,
Osaka University
2-1, Yamada-oka, Suita, Osaka 565, Japan
  
Compiler of data for Sinbad:
S. Kitsos
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
  
Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia
  
Reviewer of compiled data:
A. Trkov
IAEA/Nuclear Data Section                
Wagramerstrasse 5, P.O.Box 100, A-1400 Vienna, Austria
  
I. Kodeli
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France

NEA-1553/53
SINBAD-OKTAVIAN-W
=================
Author/Organizer:
----------------
Chihiro Ichihara, Katsuhei Kobayashi:
Research Reactor Institute, Kyoto University
Noda, Sennan-gun, Osaka 590-04, Japan
  
Shu A. Hayashi:
Institute for Atomic Energy, Rikkyo University
2-5-1 Nagasaka, Yokosukas Kanagawa 240-01, Japan
  
Itsuro Kimura:
Department of Nuclear engineering, Faculty of Engineering,
Kyoto University
Yoshida-honmachi, Sakyo-ku, Kyoto 606, Japan
  
Junji Yamamoto, Akito Takahashi, T. Kanaoka, I. Murata, K. Sumita:
Department of Nuclear Engineering, Faculty of Engineering,
Osaka University
2-1, Yamada-oka, Suita, Osaka 565, Japan
  
Compiler of data for Sinbad:
S. Kitsos
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
  
Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia
  
Reviewer of compiled data:
I. Kodeli
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France

NEA-1553/54
SINBAD-FNG-BLKT
===============
Author/Organizer
----------------
Experiment and analysis:
P. Batistoni, M. Angelone, M. Pillon, L. Petrizzi, Associazione Euratom -
ENEA sulla Fusione, C. R. Frascati - I- C. P. 65, 00044 Frascati, Italy

Compiler of data for SINBAD:
Hamilton Hunter, Radiation Shielding Information Center, Oak Ridge National
Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6362, fax 423-574-6182

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Reviewer of compiled data:
P. Batistoni, ENEA Fusion Division Via E. Fermi 27, I-00044 Frascati
(Roma), ITALY

NEA-1553/55
SINBAD-FNG-DOSE-RATE
====================
Author/Organizer:
----------------
Experiment and analysis:
P. Batistoni, M. Angelone, M. Pillon, L. Petrizzi
ENEA
Centro Ricerche Energie Frascati
UTS Fusione
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy

Compiler of data for Sinbad:
P. Batistoni
ENEA
Centro Ricerche Energie Frascati
UTS Fusione
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

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

Acknowledgement
---------------
The experiment and the corresponding analysis was performed in the framework of
the EFDA (European Fusion Development Agreement) ITER Task (T-426-1998/2000).

NEA-1553/56
SINBAD-FNG-SIC
==============
Author/Organizer:
----------------
Experiment and analysis:
P. Batistoni, M. Angelone, M. Pillon, L. Petrizzi
ENEA
Centro Ricerche Energie Frascati
UTS Fusione
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy

Compiler of data for Sinbad:
P. Batistoni
ENEA
Centro Ricerche Energie Frascati
UTS Fusione
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

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


Acknowledgement
---------------
The experiment and the corresponding analysis was performed in the framework of the
EFDA (European Fusion Development Agreement) Task (TTMN-002-2001).

NEA-1553/57
SINBAD-FNG-SS
=============
Author/Organizer:
----------------
Experiment and analysis:
M. Martone, M. Angelone,  P. Batistoni,  M. Pillon, V. Rado
Neutronics Division - Fusion Department
ENEA - Ente per le Nuove tecnologie, l'Energia e l'Ambiente
C. R. Frascati - I - 00044 FRASCATI (Italy)

Compiler of data for SINBAD:
Hamilton Hunter, Radiation Shielding Information Center, Oak Ridge National
Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6362, fax 423-574-6182,

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Reviewer of compiled data:
P. Batistoni, ENEA Fusion Division Via E. Fermi 27, I-00044 Frascati
(Roma), ITALY

NEA-1553/58
SINBAD-FNG-STREAMING
====================
Author/Organizer:
----------------
Experiment and analysis:
P. Batistoni, M. Angelone, M. Pillon, L. Petrizzi
ENEA
Centro Ricerche Energie Frascati
UTS Fusione
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy

Compiler of data for Sinbad:
P. Batistoni
ENEA
Centro Ricerche Energie Frascati
UTS Fusione
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

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


Acknowledgement
---------------
The experiment and the corresponding analysis was performed in the framework of the
EFDA (European Fusion Development Agreement) ITER Task (T-362-1997).

NEA-1553/59
SINBAD-FNG-W
============
Author/Organizer:
----------------
Experiment and analysis:

P. Batistoni, M. Angelone, M. Pillon, L. Petrizzi
ENEA
Centro Ricerche Energie Frascati
UTS Fusione
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy

Compiler of data for Sinbad:
P. Batistoni
ENEA
Centro Ricerche Energie Frascati
UTS Fusione
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

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


Acknowledgement
---------------
The experiment and the corresponding analysis was performed in the framework of the EFDA (European Fusion Development Agreement) Task (TTMN-002-2002).

NEA-1553/60
SINBAD-FNS-DUCT
===============
Author/Organizer
----------------
Experiment and analysis:
Yuichi MORIMOTO, Kentaro OCHIAI, Takashi NISHIO, Masayuki WADA,
Michinori YAMAUCHI(*) and Takeo NISHITANI(**)
*Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun,
Ibaraki-ken 319-1195, Japan

Compiler of data for Sinbad:
S. Kitsos
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

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

NEA-1553/61
SINBAD-FNS-OXYGEN
=================
Author/Organizer:
----------------
Experiment and analysis:
Yukio OYAMA and Hiroshi MAEKAWA
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 Japan
Phone: 81-292-82-6809 (for Y.O.) or -6015 (for H.M.)
Fax: 81-292-82-5996 (for Y.O.) or -6365 (for H.M.)

Compiler of data for Sinbad:
I. Kodeli,
Institute Jozef Stefan, Jamova 39, 1000 Ljubljana, Slovenia

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Reviewer of compiled data:
A. Trkov
Institute Jozef Stefan, Jamova 39, 1000 Ljubljana, Slovenia

F. Maekawa
JAERI, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195  JAPAN

NEA-1553/62
SINBAD-FNS-SKYSHINE
===================
Author/Organizer:
----------------
Experiment and analysis:
T. Nishitani, K. Ochiai, J. Hori, S. Sato, M. Yamauchi, M. Nakao, M. Wada, C. Kutukake, S. Tanaka
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken 319-1195, Japan
Phone: +81-29-282-6859 (for T.N)
FAX: +81-29-282-5709 (for T.N)

S. Yoshida, R. Tanaka, I. Murata, A. Takahashi
Osaka University
Yamadaoka 2-1, Suita, Osaka, 565-0871, Japan

J. Kaneko, M. Wakisaka, T. Sawamura
Hokkaido University
Kita 8 Nishi 5, Kita-ku, Sapporo 060-0808, Japan

Compiler of data for Sinbad:
S. Kitsos
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

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

The data were kindly contributed by Dr. Takeo Nishitani.

NEA-1553/69
SINBAD-TUD-FNG-BS
=================

Author/Organizer:
----------------
Experiment and analysis:
H. Freiesleben, D. Richter, K. Seidel, S. Unholzer
Technische Universitaet Dresden
Institut fuer Kern- und Teilchenphysik
D-01062 Dresden
Germany

W. Hansen
Technische Universitaet Dresden
Institut fuer Energietechnik
D-01062 Dresden
Germany

U. Fischer, Y. Wu
Forschungszentrum Karlsruhe
Institut fuer Kern- und Energietechnik
P.O. Box 3640
D-76021 Karlsruhe
Germany

M. Angelone, P. Batistoni, M. Pillon
ENEA
Centro Ricerche Energie Frascati
Settore Fusione - Divisione Neutronica
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy

Compiler of data for Sinbad:
K. Seidel
Technische Universitaet Dresden
Institut fuer Kern- und Teilchenphysik
D-01062 Dresden
Germany

Quality assessment:
A. Milocco,
Institut Jozef Stefan,
Jamova 39, Ljubljana, Slovenia

Reviewer of compiled data:
I. Kodeli, A. Trkov
Institute Jozef Stefan,
Jamova 39,
1000 Ljubljana, Slovenia

NEA-1553/70
SINBAD-TUD-FNG-SIC
==================
Authors/Organizer:
------------------
Experiment and analysis:
K. Seidel, H. Freiesleben, C. Negoita, S. Unholzer
Technische Universitaet Dresden
Institut fuer Kern- und Teilchenphysik
D-01062 Dresden
Germany

U. Fischer, Y. Chen, R. L. Perel
Forschungszentrum Karlsruhe
Institut fuer Reaktorsicherheit
D-76021 Karlsruhe
Germany

M. Angelone, P. Batistoni, M. Pillon
Associazione ENEA-EURATOM
Settore Fusione - Divisione Neutronica
Via E. Fermi 27
I-00044 Frascati (Rome)
Italy

Compiler of data for Sinbad:
K. Seidel
Technische Universitaet Dresden
Institut fuer Kern- und Teilchenphysik
D-01062 Dresden, Germany

Quality assessment:
A. Milocco,
Institut Jozef Stefan,
Jamova 39, Ljubljana, Slovenia

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

NEA-1553/71
SINBAD-FNG-HCPB
===============
Author/Organizer:
----------------
Experiment and analysis:
P. Batistoni, M. Angelone, M. Pillon, L. Petrizzi
ENEA
Centro Ricerche Energie Frascati
UTS Fusione
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy
  
Compiler of data for Sinbad:
I. Kodeli
Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia
  
Quality assessment:
I. Kodeli
Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia
  
Reviewer of compiled data:
S. Villari
ENEA
Centro Ricerche Energie Frascati
UTS Fusione  
Via E. Fermi 27
C.P. 65
I-00044 Frascati (Rome)
Italy  
  
Acknowledgement
---------------
The experiment and the corresponding analysis was performed in the framework of the EFDA (European Fusion Development Agreement) Task (TTMN-002-2002).

NEA-1553/73
SINBAD-FNS-V
============
Author/Organizer:
-------------
Experiment and analysis:
F. Maekawa, C. Konno, Y. Kasugai, Y. Oyama, Y. Ikeda
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 Japan


Compiler of data for Sinbad:
I. Kodeli, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Reviewer of compiled data:
P. ORTEGO
Sea Shielding Engineering And Analysis (SEA)
Avda. Atenas 75 Local 9
Las Rozas 28230 MADRID

NEA-1553/74
SINBAD-FNS-W
============
Author/Organizer:
-------------
Experiment and analysis:
F. Maekawa, C. Konno, Y. Kasugai, Y. Oyama, Y. Ikeda
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 Japan

Compiler of data for Sinbad:
S. Kitsos
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Reviewer of compiled data:
I. Kodeli, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

NEA-1553/75
SINBAD-IPPE-FE
==============
Experiment and Analysis:
S.P. Simakov, B.V. Devkin, M.G. Kobozev, V.A. Talalaev (Inst. of Physics and Power Engineering, Obninsk), U. Fischer, U. von Moellendorff (Forschungszentrum Karlsruhe).

Quality assessment
A. Milocco
Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Reviewer of Compiled Data
I. Kodeli
OECD/NEA, 12 bd. des Iles,
92130 Issy les Moulineaux, France

NEA-1553/76
SINBAD-IPPE-
============
Experiment and Analysis:
S.P. Simakov, B.V. Devkin, B.I. Fursov, M.G. Kobozev, V.A. Talalaiev (Inst. of Physics and Power Engineering, Obninsk), U. von Moellendorff (Forschung-zentrum Karlsruhe), M. M. Potapenko (Sci. & Res. Ins. of Organic Materials).

Compiler of data for Sinbad:
P. Ortego
SEA, Shielding Engineering and Analysis S.L.
Avda. Atenas 75 Las Rozas, 28230 Madrid, Spain
   

Quality assessment:
A. Milocco
Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Reviewer of Compiled Data:
I. Kodeli
OECD/NEA, 12 bd. des Iles,
92130 Issy les Moulineaux, France

The data were contributed by Dr. S.P. Simakov.

NEA-1553/77
SINBAD-OKTAVIAN/AL
=================
Author/Organizer:
-----------------
Experiment and analysis:
Chihiro Ichihara, Katsuhei Kobayashi,
Research Reactor Institute, Kyoto University
Noda, Sennan-gun, Osaka 590-04, Japan

Shu A. Hayashi,
Institute for Atomic Energy, Rikkyo University
2-5-1 Nagasaka, Yokosukas Kanagawa 240-01, Japan

Itsuro Kimura,
Department of Nuclear engineering, Faculty of Engineering,
Kyoto University
Yoshida-honmachi, Sakyo-ku, Kyoto 606, Japan

Junji Yamamoto, Akito Takahashi, T. Kanaoka, I. Murata, K. Sumita,
Department of Nuclear Engineering, Faculty of Engineering,
Osaka University
2-1, Yamada-oka, Suita, Osaka 565, Japan

Compiler of data for Sinbad:
A. Trkov,
Institute Jozef Stefan, Jamova 39, 1000 Ljubljana, Slovenia

Quality assessment:
A. Milocco, Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Reviewer of compiled data:
I. Kodeli
Institute Jozef Stefan, Jamova 39, 1000 Ljubljana, Slovenia

F. Maekawa
JAERI, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195  JAPAN

NEA-1553/78
SINBAD-OKTAVIAN-W
=================
Author/Organize
--------------
- Chihiro Ichihara, Katsuhei Kobayashi:
Research Reactor Institute, Kyoto University
Noda, Sennan-gun, Osaka 590-04, Japan
- Shu A. Hayashi:
Institute for Atomic Energy, Rikkyo University
2-5-1 Nagasaka, Yokosukas Kanagawa 240-01, Japan
- Itsuro Kimura:
Department of Nuclear engineering, Faculty of Engineering,
Kyoto University
Yoshida-honmachi, Sakyo-ku, Kyoto 606, Japan
- Junji Yamamoto, Akito Takahashi, T. Kanaoka, I. Murata, K. Sumita:    Department of Nuclear Engineering, Faculty of Engineering,
Osaka University
2-1, Yamada-oka, Suita, Osaka 565, Japan

Compiler of data for Sinbad:
A. Milocco
Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Quality assessment:
A. Milocco
Institut Jozef Stefan, Jamova 39, Ljubljana, Slovenia

Reviewer of compiled data:
I. Kodeli
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
top ]
16. MATERIAL AVAILABLE
NEA-1553/01
SINBAD FUSION FULL SET

NEA-1553/26
UIFE_A.HTM Abstract
UIFE_E.HTM Description of experiment
CFSPEC~1.TXT
CFSPECE.TXT
UIFE_E1.GIF Physical Arrangement for the iron sphere
UIFE_E2.GIF Background Contributions Layout
UIFE_E3.GIF Leakage Spectra for Cf-252 source in iron sphere
UIFE_E4.GIF Leakage Spectra for D-T source in iron sphere
CFSPECN.TXT
References

NEA-1553/27
Figure 1 - SB5 Experimental Facility
Figure 2 - Calculated Neutron Source Spectrum
Figure 3 - Unfolded Spectrum for T(d,n)He-4
Figure 4 - Target Holder
Figure 5 - Experimental Geometry
Figure 6 - Additional Schematics of Experiment
Figure 7 - Three-Dimensional View of Experimental Enclosure
Figure 8 - Two-Dimensional Calculational Model
DET1.GIF NE-213 Detector
Figure 10 - Detector Locations and Shield Configurations
SB5NSP.GIF Unfolded neutron spectra
SB5TABLE.HTM Tables 1-66 Neutron and Gamma-ray Results
SRCFORT.HTM FORTRAN Source File
SB5_A.HTM Abstract
SB5_E.HTM Description of experiment
NERESG.GIF
NERESN.GIF
References

NEA-1553/40
MEPhIstr-a.htm  This information file
MEPhIstr.htm    Description of Experiment
mcnp1Y.inp      Model for Y axis MCNP-4c2 calculations of 1 mock-up
mcnp1Z.inp      Model for Z MCNP-4c2 calculations of 1 mock-up
mcnp2Z.inp      Model for Z MCNP-4c2 calculations of 2 mock-up
mcnp3Z.inp      Model for Z MCNP-4c2 calculations of 3 mock-up
mcnp4Z.inp      Model for Z MCNP-4c2 calculations of 4 mock-up
mcnp5Z.inp      Model for Z MCNP-4c2 calculations of 5 mock-up
mcnp6Z.inp      Model for Z MCNP-4c2 calculations of 6 mock-up
mcnp7Z.inp      Model for Z MCNP-4c2 calculations of 7 mock-up
mcnp8Z.inp      Model for Z MCNP-4c2 calculations
spectra.inp     Model for MCNP-4c2 spectra calculation
fig1.gif        The angular and energy dependence of the source neutron output
fig2.gif        The geometry of target units
fig3.gif        General view on mock-up
fig4.gif        View on mock-up
fig5.gif        View on mock-up
table1.xls      The angular and energy dependence of the source neutron output
table2.xls      The material of shielding composition
table3.xls      Set of threshold nuclear reactions
table4.xls      X detectors co-ordinate
table5.xls      Y detectors co-ordinate
table6-14.xls   Reaction rates in mock-ups
table15.xls     Thermoluminescent detectors
table16-18.xls  Heating rates in mock-ups
table19.xls     Basic neutron and photon spectra
table20.xls     Basic neutron and photon spectra behind the model
table21.xls     Spectra of neutrons in different X coordinates mock-up 1
table22.xls     Spectra of photons in different X coordinates mock-up 1
table23.xls     Error source of absolute value of neutron yield
table24-26.xls  Error source components
MEPhI-str1.pdf  Fusion Engineering and Design, 42, 1998, 261-266
MEPhI-str2.pdf  Fusion Engineering and Design, 55, 2001, 373-385

NEA-1553/41
fzk-be_a.htm    Abstract
fzk-be_e.htm    Description of experiment
kantmcnp.i      MCNP input for 17 cm thick shell
fzkbe.tb1       Tabulated spectrum of 5 cm shell
fzkbe.tb2       Tabulated spectrum of 10 cm shell
fzkbe.tb3       Tabulated spectrum of 17 cm shell
kant.xls        Measured spectra from Be shells
fzk-be.pdf      Reference [M95]

NEA-1553/43
Tab. (1) Properties of the activation detectors used in the model blanket
experiment.
Tab. (2) Measured saturation activities around neutron generator target(R=0.100
m), on the inner blanket surface (R=0.100 m) and in the blanket (R=0.308 m) [2].

Tab. (3) Measured saturation activities in the blanket(R=0.528 m), on the inner
Be multiplier surface (R=0.055 m) and on the inner blanket surface(R=0.100 m)-
Be multiplier present [2].
Tab. (4) Measured saturation activities in the blanket(R=0.149 m, R=0.308 m
R=0.528 m) - Be multiplier present.
Tab. (5) Measured saturation activities in the Carbon reflector(R=0.689 m) - Be
multiplier not present and in the Carbon reflector(R=0.689 m) - Be multiplier
present.
Tabs. (6,6a,6b) Measured with use of LSC and radiochemical methods and
calculated tritium production reaction rates in natural Li (Li configuration)
Tabs. (7,7a) Measured with use of LSC and calculated tritium production reaction

rates in 6Li and 7Li (Li configuration)
Tabs. (8,8a) Measured with use of LSC and calculated tritium production reaction

rates in Li blanket with graphite reflector
Tabs. (9,9a) Measured with use of LSC and calculated tritium production reaction

rates in Li blanket with inner Be multiplier
Tab. (10) Measured with use of LSC and calculated tritium production reaction
rates in Li blanket with inner Be multiplier and graphite reflector
Tab. (11) Saturation activities calculated in the simplified blanket model (Li
configuration) with the use of MCNP and FISPACT.
Tab. (12) Saturation activities calculated in the simplified blanket model
(Be-Li configuration) with the use of MCNP and FISPACT.
Tab. (13) Saturation activities calculated in the simplified blanket model (Li-C

configuration) with the use of MCNP and FISPACT.
Tab. (14) Saturation activities calculated in the simplified blanket model
(Be-Li-C configuration) with the use of MCNP and FISPACT.
Figure 1 - Li Blanket Sketch
Figure 2 - The Li blanket layout.
Figure 3 - The Li blanket layout with accelerator.
Figure 4 - The Li blanket layout with graphite reflector.
Figure 5  - Inner berylium multiplier sketch
Figure 6 - Target support sketch
Figure 7 - MCNP input

NEA-1553/45
M94-38 directory:
T221.TXT   Source neutron spectrum
T222.TXT   Source gamma-ray spectrum
T3112.TXT   Neutron Spectrum in MeV at -10, 76 and 228 mm
T3113.TXT   Neutron Spectrum in MeV at 380, 532 and 618 mm
T3121.TXT   Neutron Spectrum in keV at -10, 76 and 228 mm
T3122.TXT   Neutron Spectrum in keV at 380, 532 and 618 mm
T322.TXT   Reaction Rate
T3311.TXT   Prompt Gamma-Ray Spectrum at  76 mm
T3312.TXT   Prompt Gamma-Ray Spectrum at 228 mm
T3313.TXT   Prompt Gamma-Ray Spectrum at 380 mm
T3314.TXT   Prompt Gamma-Ray Spectrum at 532 mm
T3322.TXT   Decay Gamma-Ray Spectrum #1
T3323.TXT   Decay Gamma-Ray Spectrum #2
T3324.TXT   Decay Gamma-Ray Spectrum #3
T3325.TXT   Decay Gamma-Ray Spectrum #4
T3326.TXT   Energy-Integrated Decay Gamma-Ray Spectrum
T343.TXT   Gamma-Ray Heating Rate
F-A-1.TXT   Input data for GRTUNCL
F-A-2.TXT   Input data for DOT3.5
F-A-3.TXT   Input data for MCNP with Neutron Source
F-A-4.TXT   Input data for MCNP with Gamma-Ray Source
IRON directory:
T-1.TXT      Neutron Spectrum in Energy Region of keV
T-2.TXT      Neutron Spectrum in Energy Region of eV
T-3.TXT      Gamma-Ray Heating Rate
NSOURCE.TXT   Source neutron spectrum
GSOURCE.TXT   Source gamma-ray spectrum
GRTUNCL.TXT   Input data for GRTUNCL
DOT35.TXT   Input data for DOT3.5
MCNP.TXT   Input data for MCNP
M94-14 directory: Tables extracted from JAERI-M 94-014
JAERI-M 94-014, JAERI-M 94-015 and JAERI-M 94-038 documents

NEA-1553/46
tufe-abs.htm  Information file
tufe-exp.htm Description of Experiment
Quality assessment of FNG and TUD benchmarks (report IJS-DP-10216)
MCNP.DAT  3-D model for MCNP-4A calculations (high quality)
FIG-1.TIF  Angular dependence of the source intensity (high quality)
FIG-2.TIF  Angular dependence of the source energy distribution
FIG-3.TIF  Geometries A0, A1, and A2 (horizontal section)
FIG-4.TIF  Geometries A0, A1, and A2 (vertical section)
FIG-5.TIF  Neutron spectra (high quality)
FIG-6.TIF  Neutron time of arrival spectra (high quality)
FIG-7.TIF  Photon spectra (high quality)
FIG-8.TIF  Neutron detector efficiency of the NE213 detector
FIG-9.TIF  Neutron spectra for A0 geometry (experiment/MCNP)
FIG-10.TIF Neutron time of arrival spectra for A0 geometry
FIG-11.TIF Photon spectra for A0 geometry (experiment/MCNP)
FIG-1.GIF  Angular dependence of the source intensity (preview)
FIG-2.GIF  Angular dependence of the source energy distribution
FIG-3.GIF  Geometries A0, A1, and A2 (horizontal section)
FIG-4.GIF  Geometries A0, A1, and A2 (vertical section)
FIG-5.GIF  Neutron spectra
FIG-6.GIF  Neutron time of arrival spectra
FIG-7.GIF  Photon spectra
FIG-8.GIF  Neutron detector efficiency of the NE213 detector
FIG-9.GIF  Neutron spectra for A0 geometry (experiment/MCNP)
FIG-10.GIF Neutron time of arrival spectra for A0 geometry
FIG-11.TIF Photon spectra for A0 geometry (experiment/MCNP)

NEA-1553/47
tudw-a.htm    abstract
tudw-e.htm    Description of experiment
tudw-c.htm    Description of transport calculations
FNG-TUD.pdf   Document describing the quality assessment of the FNG and TUD
benchmarks (reference 8)
D-T.pdf       Document describing the D-T source routine for MCNPX(5) (ref.9)
DT_MCNP5.TXT  Patch with MCNP5 source subroutines for the calculation of 14-MeV
D-T source (new release)
source.F      source.F subroutine for MCNPX-2.6f to calculate 14-MeV D-T source
(new release)
srcdx.F       srcdx.F subroutine for MCNPX-2.6f containing also subroutines for
numerics  to calculate 14-MeV D-T source (new revised version)
source.for    FNG D-T source routine for MCNP4C (obsolete)
mcnp.inp      3-D model for MCNP-4C code
fig1.jpg      Fig. 1: Angular dependence of the neutron source
fig2.jpg      Fig. 2: Neutron source energy/angular distribution
fig3.jpg      Fig. 3: Geometry of the assembly
fig4.jpg      Fig. 4: Neutron spectra at position P-1
fig5.jpg      Fig. 5: Neutron spectra at position P-2
fig6.jpg      Fig. 6: Neutron spectra at position P-3
fig7.jpg      Fig. 7: Neutron spectra at position P-4
fig8.jpg      Fig. 8: Gamma-ray spectra at position P-1
fig9.jpg      Fig. 9: Gamma-ray spectra at position P-2
fig10.jpg     Fig. 10: Gamma-ray spectra at position P-3
fig11.jpg     Fig. 11: Gamma-ray spectra at position P-4
eff-857.pdf   Reference [6] EFFDOC-857
eff-860.pdf   Reference [7] EFFDOC-860

NEA-1553/48
okni-abs.htm  Abstract
okni-exp.htm  Description of Experiment
Oktavian.pdf  Document describing quality assessment of OKTAVIAN experiments
NI2d.i  Recommended 2-D MCNP5(X) input allowing time and energy domain analysis
mcnp1d.inp    1-D MCNP-4B Input (OBSOLETE-cavity & detector modeling
approximations)
mcnp3d.inp    3-D MCNP-4B Input (OBSOLETE-cavity & detector modeling
approximations)
Ni-Fig1.tif   Plan view of the OKTAVIAN facility (high quality)
Ni-Fig2.tif   View of the experimental arrangement (high quality)
Ni-Fig3.tif   Angular dependence of the neutron source energy (high quality)
Ni-Fig4.tif   Angular dependence of the neutron yield (high quality)
Ni-Fig5.tif   Leakage neutron spectrum (high quality)
Ni-Fig6.tif   Leakage neutron spectrum (high quality)
Ni-Fig7.tif   Source spectra comparison (high quality)
Ni-Fig8.tif   Calculated leakage spectrum dependence on the source
              spectrum (high quality)
Ni-Fig9.tif   Calc.leakage spectrum dependence on the geometrical
              model (1D/3D) (high quality)
Ni-F10a.tif   Calc.spectrum dependence on nuclear data
              (high energy)-3D model (high quality)
Ni-F10b.tif   Calc.spectrum dependence on nuclear data
              (low energy)-3D model (high quality)
Ni-Fig1.gif   Plan view of the OKTAVIAN facility (preview)
Ni-Fig2.gif   View of the experimental arrangement (preview)
Ni-Fig3.gif   Angular dependence of the neutron source energy (preview)
Ni-Fig4.gif   Angular dependence of the neutron yield (preview)
Ni-Fig5.gif   Leakage neutron spectrum
Ni-Fig6.gif   Leakage neutron spectrum
Ni-Fig7.gif   Source spectra comparison
Ni-Fig8.gif   Calculated leakage spectrum dependence on the source spectrum
Ni-Fig9.gif   Calc.leakage spectrum dependence on the geometrical model (1D/3D)
Ni-F10a.gif   Calc.spectrum dependence on nuclear data (high energy)-3D model
Ni-F10b.gif   Calc.spectrum dependence on nuclear data (low energy)-3D model
OKTNI_2.pdf   Reference
OKTNI_1.pdf   Reference
OKTNI_3.pdf   Reference
ane-10.pdf    Reference

NEA-1553/50
okfe-abs.htm  Abstract
okfe-exp.htm  Description of Experiment
Oktavian.pdf  Document describing quality assessment of OKTAVIAN experiments
FE2d.i        Recommended 2-D MCNP5(X) input allowing time and energy domain
analysis
Mcnp1.inp     1-D MCNP-4B Input by F. Maekawa (OBSOLETE-cavity & detector
modeling approximations)
Mcnp2.inp     3-D MCNP-4B Input by A. Trkov (OBSOLETE-cavity & detector modeling

approximations)
Fe-Fig1.tif   Figure 1: Plan view of the OKTAVIAN facility (high quality)
Fe-Fig2.tif   Figure 2: View of the experimental arrangement (high quality)
Fe-Fig3.tif   Figure 3: Iron sphere details (high quality)
Fe-Fig1.gif   Figure 1: Plan view of the OKTAVIAN facility (preview)
Fe-Fig2.gif   Figure 2: View of the experimental arrangement (preview)
Fe-Fig3.gif   Figure 3: Iron sphere details (preview)
Fe-Fig4.gif   Figure 4: Angular dependence of the neutron source (preview)
Fe-Fig5.gif   Figure 5: Angular dependence of the neutron yield (preview)
Fe-Fig6.gif   Figure 6: Leakage neutron spectrum (preview)
Fe-Fig7.gif   Calculated vs. experimental neutron spectra - JENDL-FF (F.
Maekawa- 1D model)
Fe-Fig8.gif   Calculated vs. experimental neutron spectra - EFF-3.1 (A. Trkov-
1D model)
OKTFE_1.pdf, Oktfe_5.pdf, Oktfe_8.pdf, ane-10.pdf   References

NEA-1553/52
oksi-abs.htm   Abstract
oksi-exp.htm   Description of experiment
Oktavian.pdf   Document describing quality assessment of OKTAVIAN experiments
SI42dn.i       Routine (~2D) MCNP5(X) input with neutron source (Si 40cm)
SI42dg.i       Routine (~2D) MCNP5(X) input with gamma source (Si 40cm)
SI43dn.i       Detailed 3D MCNP5(X) input for neutron spectrum calc. (Si 40cm)
SI43dg.i       Detailed 3D MCNP5(X) input with DT source -gamma calc. (Si 40cm)
SI62dn.i       Routine (~2D) MCNP5(X) input with neutron source (Si 40cm)
SI62dg.i       Routine (~2D) MCNP5(X) input with gamma source (Si 40cm)
SI63dn.i       Detailed 3D MCNP5(X) input for neutron spectrum calc. (Si 40cm)
SI63dg.i       Detailed 3D MCNP5(X) input with DT source -gamma calc. (Si 40cm)
DT_MCNP5.TXT   patch with the source subroutines for MCNP5
               to calculate 14-MeV D-T source (new revised version)
source.F       source.F subroutine for MCNPX version 2.6f
               to calculate 14-MeV D-T source (new revised version)
srcdx.F        srcdx.F subroutine for MCNPX version 2.6f
               containing also subroutines for numerics
               to calculate 14-MeV D-T source (new revised version)
D-T.pdf        Document describing D-T source routine for MCNPX/MCNP5
mcnp60_n.inp   1D Input for MCNP4B neutron calculations - Si 60cm (OBSOLETE)
mcnp60_g.inp   1D Input for MCNP4B gamma calculations - Si 60 cm (OBSOLETE)
oksi-f1.gif    Figure 1: Experimental arrangement of the OKTAVIAN Facility
oksi-f2.gif    Figure 2: 60 cm diameter vessel (Type-III)
oksi-f3.gif    Figure 3: 40 cm diameter vessel (Type-II)
oksi-f4.gif    Figure 4: Gamma-ray emission spectrum from the neutron source
oksi-f5.jpg    Figure 5: Gamma-ray source spectrum from 1D MCNP input and Table
2
oksi-f6.gif    Figure 6: Neutron spectra from Si 60 cm pile, ref. [8]
oksi-f7.gif    Figure 7: Gamma-ray spectra from Si 40 cm pile, ref. [8]
oksi-f8.gif    Figure 8: Gamma-ray spectra from Si 60 cm pile, ref. [8]
oksi-f9.gif    Figure 9: Neutron spectra from Si 40 cm pile - ENDF/B-VI, ref.
[9]
oksi-f10.gif   Figure 10: Neutron spectra from Si 60 cm pile, ref. [9]
oksi-f11.jpg   Figure 11: Neutron spectra from Si 60 cm pile - EFF-3, ENDF/B-VI
(by A. Trkov)
oksi-f12.gif   Figure 12: Gamma-ray spectra from a Si 60 cm pile, ref. [9]
oksi-f13.gif   Figure 13: Gamma-ray spectra from a Si 60 cm pile - ENDF/B-VI,
ref. [9]
j94-014.pdf    Reference
j98-024.pdf    Reference
ane-10.pdf     Reference
si40-n-exp.txt Low quality neutron leakage spectrum for the Si-40cm shell
experiment

NEA-1553/53
okw-abs.htm   Abstract
okw-exp.htm   Description of experiment
Oktavian.pdf  Document describing quality assessment of OKTAVIAN experiments
W2dns.i       Routine (~2D) MCNP5(X) input with neutron source
W2dgs.i       Routine (~2D) MCNP5(X) input with gamma source
W3dn.i        Detailed 3D MCNP5(X) input for neutron spectrum analysis
W3dg.i        Detailed 3D MCNP5(X) input with DT source -gamma analysis
DT_MCNP5.TXT  patch with the source subroutines for MCNP5
              to calculate 14-MeV D-T source (new revised version)
source.F      source.F subroutine for MCNPX version 2.6f
              to calculate 14-MeV D-T source (new revised version)
srcdx.F       srcdx.F subroutine for MCNPX version 2.6f
              containing also subroutines for numerics
              to calculate 14-MeV D-T source (new revised version)
D-T.pdf       Document describing D-T source routine for MCNPX/MCNP5
mcnp4b-n.inp  MCNP4B input for neutron calculation (W 40 cm) (OBSOLETE)
mcnp4b-g.inp  MCNP4B input for gamma calculation (W 40 cm) (OBSOLETE)
okw-f1.gif    Fig. 1: Experimental arrangement of the OKTAVIAN Facility
okw-f2.gif    Fig. 2: 40 cm diameter vessel (Type-II)
okw-f3.jpg    Fig. 3: Comparison of neutron source spectrum from MCNP
                      input and Table 1
okw-f4.gif    Fig. 4: Gamma-ray emission spectrum from neutron source
okw-f5.jpg    Fig. 5: Comparison of gamma-ray source spectrum from MCNP
                      input and Table 2
okw-f6.jpg    Fig. 6: Measured and calculated neutron spectra (lethargy
                      scale) (from ref. [10])
okw-f7.jpg    Fig. 7: Measured and calculated neutron spectra (energy
                      scale) [10]
okw-f8.jpg    Fig. 8: Measured and calculated gamma-ray spectra [10]
j94-014.pdf   Reference
j98-024.pdf   Reference
ane-10.pdf    Reference

NEA-1553/54
FNGBKT_a.HTM   Abstract
FNGBKT_e.HTM   Description of Experiment
FNG-TUD.pdf    Document describing the quality assessment of FNG and TUD
benchmarks
DETEC.DAT      List of activation reactions, foil locations, and foil size
RREXP.DAT      Experimental data for reaction rate measurements.
RRCALC.DAT     Calculated reaction rates using FENDL-1 library and MCNP
NHEXP.DAT      Expermental dose measurements for nuclear heating.
NHCALC.DAT     Calculated nuclear heating using MCNP/FENDL-1/EFF-3
SOURCE.DAT     Source neutron spectrum averaged over spherical cap of 60 degree
aperture.
DT_MCNP5.TXT   patch with the source subroutines for MCNP5 to calculate 14-MeV
D-T source (new revised version)
source.F       source.F subroutine for MCNPX version 2.6f to calculate 14-MeV
D-T source (new revised version)
srcdx.F        srcdx.F subroutine for MCNPX version 2.6f containing also
subroutines for numerics to calculate 14-MeV D-T source (new revised version)
D-T.pdf        Document describing the D?T source routine for MCNPX/MCNP5
SOURCE.FOR     D?T FORTRAN source routine for MCNP-4A (obsolete).
BB_fast.i      Recommended MCNPX/MCNP5 input for fast reaction rate calculations

BB_Mn.i        Recommended MCNPX/MCNP5 input for Mn55(n,g) foils calculation
BB_Au.i        MCNPX/MCNP5 input for Au197(n,g) foils
MCNPGEO.TXT    Geometrical data of the experiment listed in MCNP input format
GEOM.DAT       Geometry input file used by MCNP-4A (obsolete)
FNGBLKT1.GIF   Fig. 1: The bulk shield mock-up configuration
FNGBLKT2.GIF   Fig. 2: The bulk shield mock-up assembly in front of FNG target,
without the Polyethylene shield
FNGBLKT3.GIF   Fig. 3: The bulk shield mock-up assembly in front of FNG target,
with the Polyethylene shield.
FNGBLKT4.GIF   Fig. 4: The mock-up placed on the movable tower at the centre of
FNG bunker, at 4 m from the floor, and 5.3 cm from the neutron source.
nsc_doc95.pdf  Reference 7

NEA-1553/55
fngdos-a.htm     Abstract
fngdos-e.htm     Description of Experiment
FNG-TUD.pdf      Document describing the quality assessment of the FNG and TUD
benchmarks
D-T.pdf          Document describing the D?T source routine for MCNPX(5)
DT_MCNP5.TXT     patch with the source subroutines for MCNP5 to calculate 14-MeV
D-T source (new revised version)
source.F         source.F subroutine for MCNPX version 2.6f to calculate 14-MeV
D-T source (new revised version)
srcdx.F          srcdx.F subroutine for MCNPX version 2.6f containing also
subroutines for numerics to calculate 14-MeV D-T source (new revised version)
source.for       FNG D?T source routine for MCNP4C (obsolete)
mcnp_n.inp       3-D model for MCNP-4C calculations of neutron flux (irradiation
model)
mcnp_g.inp       3-D model for MCNP-4C calculations of dose rate (decay gamma
transport, shut down model)
fisp_620.inp     Input for FISPACT run (e.g. for cell 620)
fisp_620.flx     Neutron flux, calculated in mcnp_n.inp, for FISPACT run (e.g.
for cell 620)
fisp_620.out     Output of FISPACT run (e.g. for cell 620) containing decay
g-ray spectra to be input in mcnp_g.inp
fig1.gif         Fig. 1: Angular dependence of the source
fig2.gif         Fig. 2: Energy/angular dependence of the source
fig3.gif         Fig. 3: Geometry of the experimental mock-up
fig4.gif         Fig. 4: Nickel activation foil positions in the cavity
fig5.gif         Fig. 5: Neutron irradiaton time profile
fig6.gif         Fig. 6: Percent contributions of most important radioisotopes
to the total contact dose rate, calculated by FISPACT for cell 620 of
mcnp_n.inp.
fig7.gif         Fig. 7: Measured dose rate in the cavity centre as a function
of cooling time.
fig8.gif         Fig. 8: Geometry of the TiT target
fng-dose.pdf 2,  Ref. 1
nd2001.pdf       Ref. 2
eff-726.pdf      Ref. 3
eff-614.pdf      Ref. 6

NEA-1553/56
fngsic-a.htm     Abstract
fngsic-e.htm     Description of Experiment
FNG-TUD.pdf      Document describing the quality assessment of the FNG and TUD
benchmarks
D-T.pdf          Document describing the D?T source routine for MCNPX(5)
DT_MCNP5.TXT     patch with the source subroutines for MCNP5 to calculate 14-MeV
D-T source (new revised version)
source.F         source.F subroutine for MCNPX version 2.6f to calculate 14-MeV
D-T source (new revised version)
srcdx.F          srcdx.F subroutine for MCNPX version 2.6f containing also
subroutines for numerics to calculate 14-MeV D-T source (new revised version)
source.for       FNG D?T source routine for MCNP4C (obsolete)
mcnp.inp         3-D model for MCNP-4A calculations of neutron activation
reaction rates
mcnp_tld.inp     3-D model for MCNP-4A calculations of nuclear heating
trx-sic.inp      Input data for TRANSX cross-section preparation
dort-sic.inp     Input data for GRTUNCL first collision source and DORT
transport codes
2dant-si.inp     Input data for TWODANT transport code
fig1.gif         Fig. 1: Angular dependence of the source
fig2.gif         Fig. 2: Energy/angular dependence of the source
fig3.gif         Fig. 3: Geometry of the experimental mock-up
fig4.gif         Fig. 4: Geometry of the source
fig5.gif         Fig. 5: Geometry of TLD detector
enea-sic.pdf     Reference 1
fed2002.pdf      Reference 2
soft-22.pdf      Reference 3
eff-818.pdf      Reference 4

NEA-1553/57
FNGSS_A.HTM   Abstract
FNGSSSRC.TXT  Tables 1 - 5 (see below)
FNGSS_E.HTM   Description of Experiment
FNG-TUD.pdf   Document describing the quality assessment of FNG and TUD
benchmarks
DT_MCNP5.TXT  patch with the source subroutines for MCNP5 to calculate 14-MeV
D-T source (new revised version)
source.F      source.F subroutine for MCNPX version 2.6f to calculate 14-MeV D-T
source (new revised version)
srcdx.F       srcdx.F subroutine for MCNPX version 2.6f containing also
subroutines for numerics to calculate 14-MeV D-T source (new revised version)
D-T.pdf       Document describing the D?T source routine for MCNPX/MCNP5
au.i          Recommended MCNPX(5) input for Au197(n,g) foils
au+walls.i    Recommended MCNPX(5) input for Au foils with bunker and walls
MCNPGEO.TXT   Geometry input file used by MCNP-4A (obsolete)
FNGSSF1.GIF   Fig. 1: Source, Target, and SS block diagram w/dimensions

NEA-1553/58
fngstr-a.htm   Abstract
fngstr-e.htm   Description of Experiment
FNG-TUD.pdf    Document describing the quality assessment of FNG and TUD
benchmarks
D-T.pdf        Document describing the D?T source routine for MCNPX/MCNP5
DT_MCNP5.TXT   patch with the source subroutines for MCNP5 to calculate 14-MeV
D-T source (new revised version)
source.F       source.F subroutine for MCNPX version 2.6f to calculate 14-MeV
D-T source (new revised version)
srcdx.F        srcdx.F subroutine for MCNPX version 2.6f containing also
subroutines for numerics to calculate 14-MeV D-T source (new revised version)
source.for     FNG D?T source routine for MCNP4C (obsolete)
STR_AL.i       Updated 3-D model for MCNP5(X) calculations of neutron activation
reaction rates
mcnpfoil.inp   3-D model for MCNP-4A calculations of neutron activation reaction
rates (obsolete)
mcnp_nh.inp    3-D model for MCNP-4A calculations of nuclear heating
mcnp_hss.inp   3-D simplified model for MCNP-4A calculations of nuclear heating
in stainless steel
mcnp_hcu.inp   3-D simplified model for MCNP-4A calculations of nuclear heating
in Copper
mcnp_tld.inp   3-D simplified model for MCNP-4A calculations of nuclear heating
in TLD-300
trx-fng.inp    Input data for TRANSX cross-section preparation
dort-fng.inp   Input data for GIP cross-section mixing, GRTUNCL first collision
source and DORT transport codes
fig1.gif       Fig. 1: Angular dependence of the source
fig2.gif       Fig. 2: Energy/angular dependence of the source
fig3.gif       Fig. 3: Geometry of the experimental mock-up
fig4.gif       Fig. 4: Activation foils position in the channel and the cavity
fig5.gif       Fig. 5: Geometry of the source
fed2000.pdf    Reference 4
fed2000a.pdf   Reference 5
eff-639.pdf    Reference 8
eff-673.pdf    Reference 9
eff-698.pdf    Reference 15

NEA-1553/59
fngw-a.htm     Abstract
fngw-e.htm     Description of Experiment
fngw-c.htm     Transport calculations - description & results
FNG-TUD.pdf    Document describing the quality assessment of the FNG and TUD
benchmarks
D-T.pdf        Document describing the D?T source routine for MCNPX/MCNP5
DT_MCNP5.TXT 2 patch with the source subroutines for MCNP5 to calculate 14-MeV
D-T source (new revised version)
source.F     8 source.F subroutine for MCNPX version 2.6f to calculate 14-MeV
D-T source (new revised version)
srcdx.F      9 srcdx.F subroutine for MCNPX version 2.6f containing also
subroutines for numerics to calculate 14-MeV D-T source (new revised version)
source.for     FNG D?T source routine for MCNP4C (obsolete)
FeIn.mcp       3-D model for MCNP-4C calculation of activation reaction rates
(Fe and In foils)
NbNiAu.mcp     3-D model for MCNP-4C calculation of activation reaction rates
(Nb, Ni and Au foils)
ZrAlMn.mcp     3-D model for MCNP-4C calculation of activation reaction rates
(Z, Al and Mn foils)
mcnp_tld.inp   3-D model for MCNP-4C calculation of gamma dose rates
trx-w.inp      Input data for TRANSX cross-section preparation
dort-w.inp     Input data for GRTUNCL first collision source and DORT transport
codes
2dant-w.inp    Input data for TWODANT transport code
fig1.gif       Fig. 1: Angular dependence of the source
fig2.gif       Fig. 2: Energy/angular dependence of the source
fig3.gif       Fig. 3: Geometry of FNG target
fig4.gif       Fig. 4: Y-Z view of FNG-W mock-up with detectors (X=0)
fig5.gif       Fig. 5: Y-Z view of FNG-W (X=0)
fig6.gif       Fig. 6: X-Y view of FNG-W (-3.5 cm < Z < 3.5 cm)
fig7.gif       Fig. 7: X-Y view of FNG-W (-15.5 cm < Z < -3.5 cm) and (3.5 cm cm

< Z < 15.5 cm)
fig8.gif       Fig. 8: X-Z view of the mock-up (Y=0)
fig9.gif       Fig. 9: Geometry of TLD detector
fig10.jpg      Fig. 10: Geometry model used in codes DORT and TWODANT
enea-w.pdf     Reference 1
kodeli-w.pdf   Reference 8
eff-867.pdf    Reference 9
eff-885.pdf    Reference 10
kyoto03.pdf    Reference 11

NEA-1553/60
fnsstr-a.htm    Abstract
fnsstr-e.htm    Description of the experiment
FNS.pdf         Document describing quality assessment of FNS experiments
SINBAD_DTSRC.i  Recommended MCNPX/MCNP5 input using D-T source routine
DT_MCNP5.TXT    patch with the source subroutines for MCNP5 to calculate 14-MeV
D-T source (new revised version)
source.F        source.F subroutine for MCNPX version 2.6f to calculate 14-MeV
D-T source (new revised version)
srcdx.F         srcdx.F subroutine for MCNPX version 2.6f containing also
subroutines for numerics to calculate 14-MeV D-T source (new revised version)
D-T.pdf         Document describing D?T source routine for MCNPX/MCNP5
mcnp-F2.inp     MCNP4B/4C input (FENDL/2 library)- no DT source routine
(obsolete)
mcnp-J33.inp    MCNP4B/4C input (JENDL-3.3 library)- no DT source routine
(obsolete)
fns-str-f1.gif  Fig. 1: Layout of the target room
fns-str-f2.gif  Fig. 2: Schematic view of the experimental assembly
fns-str-f3.gif  Fig. 3: Neutron spectra measured at positions #3, #5, #7 and #9
fns-str-f4.gif  Fig. 4: Reaction rates measured in the duct and behind the
assembly
fns-str-f5.gif  Fig. 5: Spectra measured and calculated with FENDL/2 at pos. #3
fns-str-f6.gif  Fig. 6: Spectra measured and calculated with FENDL/2 at pos. #5
fns-str-f7.gif  Fig. 7: Spectra measured and calculated with FENDL/2 at pos. #7
fns-str-f8.gif  Fig. 8: Spectra measured and calculated with FENDL/2 at pos. #9
fns-str-f9.gif  Fig. 9: C/E values of 93Nb(n,2n)92mNb reaction rates in the duct

and behind the assembly
fns-str-f10.gif Fig. 10: C/E values of 115In(n,n')115mIn reaction rates in the
duct and behind the assembly
fns-str-f11.gif Fig. 11: C/E values of 197Au(n,g)198Au reaction rates in the
duct and behind the assembly
fns-str-r.xls   Measured and calculated reaction rates in Microsoft Excel(tm)
format
fns-str-s.xls   Measured and calculated neutron spectra in Microsoft Excel(tm)
format
fns-str.pdf     Reference [1]

NEA-1553/61
fnso-abs.htm  Abstract
fnso-exp.htm  Description of Experiment and Results
FNS.pdf       Document describing quality assessment of FNS experiments
TOF.i         Recommended MCNPX/MCNP5 input for time domain analysis
mcnp4b.inp    Input data for MCNP-4B 3D calculations
dot35.inp     Input data for GRTUNCL/DOT-3.5 2D calculations
fnso-f1.tif   Fig. 1: Vacuum thermal insulating container geometry
fnso-f1.gif   Fig. 1: Vacuum thermal insulating container geometry
fnso-f2.gif   Fig. 2: Experimental setup for TOF Experiment
fnso-f3.gif   Fig. 3: Definition of parameters for flux calculation
fnso-f4.gif   Fig. 4: Calculational geometry for MCNP
fnso-f5.gif   Fig. 5: Measured vs. MCNP4B (ENDF/B-VI) spectra
fnso-f6.pdf   Fig. 6: Measured vs. MCNP4B (ENDF/B-VI) spectra
j90-092.pdf   Reference
j83-195.pdf   Reference
j85-031.pdf   Reference
ndst91.pdf    Reference
nse115-93.pdf Reference
j98-024.pdf   Reference
j88-101.pdf   Reference

NEA-1553/62
fnssky-a.htm    Abstract
fnssky-e.htm    Description of experiment
FNS.pdf         Document describing quality assessment of FNS experiments
mcnp4b-sky.inp  Input data for MCNP-4B calculations
mcnp4c-sky.inp  MCNP-4C Input data (including pine forest)
sky-fig1.jpg    Fig. 1: Cross-sectional view of the target room
sky-fig2.jpg    Fig. 2: Measurement points in the environment
sky-fig3.jpg    Fig. 3: Neutron dose distribution
sky-fig4.pdf    Fig. 4: Neutron spectra at 50, 100 and 200 m
sky-fig5.jpg    Fig. 5: Secondary gamma-ray spectra
sky-NaI.txt     Secondary gamma-ray pulse height spectra (numerical data)
sky-mcnp4c.xls  Measured and MCNP-4C calculated neutron dose rates
sky-ref1.pdf    Reference 1
proc_147.pdf    Reference 2

NEA-1553/69
tud-abs.htm     This information file
tud-exp.htm     Description of Experiment
FNG-TUD.pdf     Document describing the quality assessment of the
                FNG and TUD benchmarks
D-T.pdf         Document describing the D?T source routine for MCNPX(5)
DT_MCNP5.TXT    Patch with the source subroutines for MCNP5
                to calculate 14-MeV D-T source (new revised version)
source.F        source.F subroutine for MCNPX version 2.6f
                to calculate 14-MeV D-T source (new release)
srcdx.F         srcdx.F subroutine for MCNPX version 2.6f
                containing also subroutines for numerics
                to calculate 14-MeV D-T source (new release)
source.for      FNG D?T source routine for MCNP4C (obsolete)
MCNP4A.inp      3-D model for MCNP-4A calculations
TRANSX.inp      Input Data for TRANSX cross-section preparation
GIP.inp         Input Data for GIP cross-section mixing
GRTUNCL.inp     Input Data GRTUNCL first collision source code
DORT.inp        Input Data for DORT
TUD-fig1.gif    Fig. 1: Angular dependence of the source (preview)
TUD-fig2.gif    Fig. 2: Energy/angular dependence of the source (preview)
TUD-fig3.gif    Fig. 3: Geometry of the experimental mock-up (preview)
TUD-fig4.gif    Fig. 4: Neutron spectra at positions A and B (Experiment)
TUD-fig5.gif    Fig. 5: Gamma spectra at positions A and B (Experiment)
TUD-fig6.gif    Fig. 6: Neutron spectra at positions A and B (Calculation)
TUD-fig7.gif    Fig. 7: Gamma spectra at positions A and B (Calculation)
TUD-fig1.tif    Fig. 1: Angular dependence of the source (high quality)
TUD-fig2.tif    Fig. 2: Energy/angular dependence of the source (high quality)
TUD-fig3.tif    Fig. 3: Geometry of the experimental mock-up (high quality)
TUD-fig4.tif    Fig. 4: Neutron spectra at positions A and B (high quality)
TUD-fig5.tif    Fig. 5: Gamma spectra at positions A and B (high quality)
TUD-fig6.tif    Fig. 6: Neutron spectra at positions A and B (high quality)
TUD-fig7.tif    Fig. 7: Gamma spectra at positions A and B (high quality)
MCNP5(X).i      3-D model for MCNP5(X) calculations

NEA-1553/70
tudsic-a.htm   This abstract
tudsic-e.htm   Description of experiment
tudsic-c.htm   Description of transport calculations
FNG-TUD.pdf    Document describing the quality assessment of the
               FNG and TUD benchmarks
D-T.pdf        Document describing the D?T source routine for MCNPX(5)
DT_MCNP5.TXT   Patch with the source subroutines for MCNP5
               to calculate 14-MeV D-T source (new revised version)
source.F       source.F subroutine for MCNPX version 2.6f
               to calculate 14-MeV D-T source (new release)
srcdx.F        srcdx.F subroutine for MCNPX version 2.6f
               containing also subroutines for numerics
               to calculate 14-MeV D-T source (new release)
source.for     FNG D?T source routine for MCNP4C (obsolete)
mcnp.inp       3-D model for MCNP-4C code
fig1.jpg       Fig. 1: Angular dependence of the neutron source
fig2.jpg       Fig. 2: Neutron source energy/angular distribution
fig3.jpg       Fig. 3: Geometry of the assembly
fig4.jpg       Fig. 4: Neutron spectra at position P-1
fig5.jpg       Fig. 5: Neutron spectra at position P-2
fig6.jpg       Fig. 6: Neutron spectra at position P-3
fig7.jpg       Fig. 7: Neutron spectra at position P-4
fig8.jpg       Fig. 8: Gamma-ray spectra at position P-1
fig9.jpg       Fig. 9: Gamma-ray spectra at position P-2
fig10.jpg      Fig. 10: Gamma-ray spectra at position P-3
fig11.jpg      Fig. 11: Gamma-ray spectra at position P-4
eff-822.pdf    Reference [7] EFFDOC-822
eff-815.pdf    Reference [8] EFFDOC-815
soft22j39.pdf  Reference [9]
soft-22.pdf    Reference [10]

NEA-1553/71
fnghcpb-a.htm        Abstract
fnghcpb-e.htm        Description of Experiment
DT_MCNP5.TXT         Patch with the source subroutines for MCNP5
                     to calculate 14-MeV D-T source (new revised version)
source.F             source.F subroutine for MCNPX-2.6f to calculate
                     14-MeV D-T source (new revised version)
srcdx.F              srcdx.F subroutine for MCNPX-2.6f containing also
                     subroutines for numerics to calculate 14-MeV D-T
                     source (new revised version)
D-T.pdf              Document describing the D?T source routine
                     for MCNPX(5), ref.13
source.for           D?T FORTRAN source routine for MCNP-4A (obsolete).
mcnp-hcpb.i          3-D model for MCNP-4C calculation of
                     activation reaction rates (Fe and In foils)
trx-hcpb.inp         Input data for TRANSX cross-section preparation
gip-hcpb.inp         Input data for GIP cross-section preparation
dort-hcpb.inp        Input data for 2D GRTUNCL first collision source and
                     DORT transport codes
grt3-hcpb.inp        Input data for GRTUNCL3D first collision source code
tort-hcpb.inp        Input data for TORT 3D transport code
fig1.gif             Fig. 1: Angular dependence of the source
fig2.gif             Fig. 2: Energy/angular dependence of the source
fig3.gif             Fig. 3: Geometry of FNG target
hcpb-1.jpg           Fig. 4: Y-Z view of FNG-HCPB mock-up with detectors
hcpb-2.jpg           Fig. 5: X-Y view of FNG-HCPB mock-up
hcpb-3.jpg           Fig. 6: X-Z view of the mock-up (Y=0)
hcpb-4.jpg           Fig. 7: Geometry of ENEA activation foil detectors
hcpb-5.jpg           Fig. 8: Arangement of Li2CO3 pellets for T measurements
hcpb-6.jpg           Fig. 9: Geometry of JAEA detector arrangement
room-walls.jpg       Fig. 10: Geometry of room/wall around the mock-up
09-02.cdr            Figures in Corel Draw
room-walls.cdr       Figures in Corel Draw
FR_TW6_TTMN_002.pdf  Reference 1
TBM_Report4.pdf      Reference 2
effdoc-938.pdf       Reference 6
tbm-hcpbrev.pdf      Reference 7
IJS-DP-9312.pdf      Reference 11
effdoc-981.pdf       Reference 12
510_port2006.pdf     Reference 14

NEA-1553/72
fnsc-abs.htm         Abstract
fnsc-exp.htm         Description of experiment
FNS.pdf              Document describing quality assessment of FNS experiments
ACTIVATION.i         Recommended MCNP5(X) input for activation analysis
wwinp_ACTIVATION     Input for MCNP mesh-based variance reduction
N_SPECTRUM_P1.i      Recommended MCNP5(X) input for P1 neutron spectrum analysis

wwinp_N_SPECTRUM_P1  Input for MCNP mesh-based variance reduction
N_SPECTRUM_P8.i      Recommended MCNP5(X) input for P8 neutron spectrum analysis

wwinp_N_SPECTRUM_P8  Input for MCNP mesh-based variance reduction
dot35.inp            Input data for GRTUNCL and DOT calculations
mcnp4a.inp           MCNP-4A input (obsolete)
fnsc-f1.gif          Fig. 1: Sectional view of the cylindrical assembly
fnsc-f2.gif          Fig. 2: Sectional view of the graphite sheath and drawer
fnsc-f3.gif          Fig. 3: Experimental layout
fnsc-f4.gif          Fig. 4: Layout of the FNS first target room
fnsc-f5.gif          Fig. 5: Fission rates measured by micro-fission chambers
fnsc-f6.gif          Fig. 6: Foil activation measured reaction rates
fnsc-f7.gif          Fig. 7: Foil activation measured reaction rates
fnsc-f8.gif          Fig. 8: Graphite TLDs (TDL-600, 700, 100) response
fnsc-f9.gif          Fig. 9: Graphite TLDs (MSO-S, SSO-S, BSO-S) response
fnsc-f10.gif         Fig. 10: Graphite TLDs (UD-110S, 136N,137N) response
fnsc-f11.gif         Fig. 11: Fission rates measured by FDT
                     and micro-fission chambers
fnsc-f12.gif         Fig. 12: Measured neutron spectrum z=24.1 cm
fnsc-f13.gif         Fig. 13: Measured neutron spectrum z=26.7 cm
fnsc-f14.gif         Fig. 14: Measured neutron spectrum z=31.7 cm
fnsc-f15.gif         Fig. 15: Measured neutron spectrum z=41.8 cm
fnsc-f16.gif         Fig. 16: Measured neutron spectrum z=52.0 cm
fnsc-f17.gif         Fig. 17: Measured neutron spectrum z=62.1 cm
fnsc-f18.gif         Fig. 18: Measured neutron spectrum z=72.1 cm
fnsc-f19.gif         Fig. 19: Measured neutron spectrum z=77.2 cm
fnsc-f20.gif         Fig. 20: Calculation model for the graphite assembly
j88-034.pdf          Reference
j94-014.pdf          Reference
j98-024.pdf          Reference

NEA-1553/73
fnsv-abs.htm     Abstract
fnsv-exp.htm     Description of experiment
FNS.pdf          Document describing quality assessment of FNS experiments
REALISTIC.i      Recommended MCNPX/MCNP5 input using DT source routine
wwinp_REALISTIC  Input for MCNP mesh-based variance reduction
DT_MCNP5.TXT     patch with the source subroutines for MCNP5
                 to calculate 14-MeV D-T source (new revised version)
source.F         source.F subroutine for MCNPX version 2.6f
                 to calculate 14-MeV D-T source (new revised version)
srcdx.F          srcdx.F subroutine for MCNPX version 2.6f
                 containing also subroutines for numerics
                 to calculate 14-MeV D-T source (new revised version)
D-T.pdf          Document describing D?T source routine for MCNPX/MCNP5
mcnp-v.inp       MCNP-4A input taken from [1] (obsolete)
fnsv-1.gif       Fig. 1: Sectional view of the vanadium assembly
fnsv-2.gif       Fig. 2: Source neutron spectrum emitted towards 0 degree
                 from the target.
fnsv-3.gif       Fig. 3: Neutron spectra at z=7.6 cm in vanadium
fnsv-4.gif       Fig. 4: Neutron spectra at z=17.8 cm in vanadium
fnsv-5.gif       Fig. 5: Gamma-ray spectra at z=7.6 cm in vanadium
fnsv-6.gif       Fig. 6: Gamma-ray spectra at z=17.8 cm in vanadium
j98-021.pdf      Reference JAERI-Data/Code 98-021
fns-v.pdf        Reference J. Nucl. Sci. Technol., Vol. 36, No. 3

NEA-1553/74
fnsw-abs.htm     Abstract
fnsw-exp.htm     Description of experiment
FNS.pdf          Document describing quality assessment of FNS experiments
REALISTIC.i      Recommended MCNPX/MCNP5 input using DT source routine
wwinp_REALISTIC  Input for MCNP mesh-based variance reduction
DT_MCNP5.TXT     patch with the source subroutines for MCNP5
                 to calculate 14-MeV D-T source (new revised version)
source.F         source.F subroutine for MCNPX version 2.6f
                 to calculate 14-MeV D-T source (new revised version)
srcdx.F          srcdx.F subroutine for MCNPX version 2.6f
                 containing also subroutines for numerics
                 to calculate 14-MeV D-T source (new revised version)
D-T.pdf          Document describing D?T source routine for MCNPX/MCNP5
mcnp-w.inp       MCNP-4A Input taken from [2] (obsolete)
fnsw-1.gif       Fig. 1: Experimental assembly of tungsten
fnsw-2.gif       Fig. 2: Source neutron spectrum emitted towards the
                 0 degree from the target.
fnsw-3.gif       Fig. 3: Neutron spectra at z=76 mm in W (ref. [5])
fnsw-4.gif       Fig. 4: Neutron spectra at z=228 mm in W (ref. [5])
fnsw-5.gif       Fig. 5: Neutron spectra at z=380 mm in W (ref. [5])
fnsw-6.jpg       Fig. 6: Gamma-ray spectra at z=76 mm in W (ref. [5])
fnsw-7.jpg       Fig. 7: Gamma-ray spectra at z=228 mm in W (ref. [5])
fnsw-8.jpg       Fig. 8: Gamma-ray spectra at z=380 mm in W (ref. [5])
j98-021.pdf      Reference JAERI-Data/Code 98-021
j98-024.pdf      Reference JAERI-Data/Code 98-024

NEA-1553/75
ippe_fe-a.htm      Abstract
ippe_fe-e.htm      Experiment Description
mcnp_fe1.inp       MCNP-4C input for Fe sphere 1 (r= 4.5 cm, wall= 2.5 cm)
mcnp_fe2.inp       MCNP-4C input for Fe sphere 2 (r=12.0 cm, wall= 7.5 cm)
mcnp_fe3.inp       MCNP-4C input for Fe sphere 3 (r=12.0 cm, wall=10.0 cm)
mcnp_fe4.inp       MCNP-4C input for Fe sphere 4 (r=20.0 cm, wall=18.1 cm)
mcnp_fe5.inp       MCNP-4C input for Fe sphere 5 (r=30.0 cm, wall=28.0 cm)
ippefig1.jpg       Fig. 1: Experimental setup
ippefig2.jpg       Fig. 2: Configuration and sizes of five Fe spheres
ippefig3.jpg       Fig. 3: Angular/energy distribution of '14 MeV' source peak

EFF-DOC-747.pdf    EFF-DOC-747 NEA Data Bank Report
IPPE.pdf           Document describing quality assessment of IPPE experiments

NEA-1553/76
ippe_v-a.htm      Abstract
ippe_v-e.htm      Experiment Description
mcnp_r5.inp       MCNP-4C input for V sphere 1 (r=5 cm)
mcnp_r12.inp      MCNP-4C input for V sphere 2 (r=12 cm)
ippefig1.jpg      Fig. 1: Experimental setup
ippefig2.jpg      Fig. 2: 5 and 12 cm radius V spheres
ippefig3.jpg      Fig. 3: Angular/energy distribution of '14 MeV' source peak
ippefig4.jpg      Fig. 4: Uncertainties of leakage spectra from V shell 1
ippefig5.jpg      Fig. 5: Correction for time-of-flight measuring technique
indc-ccp-417.pdf  IAEA Report
fzka-6096.pdf     Karlsruhe Report
IPPE.pdf          Document describing quality assessment of IPPE experiments

NEA-1553/77
okal-abs htm    Abstract
okal-exp htm    Description of Experiment
Oktavian.pdf    Document describing quality assessment of OKTAVIAN experiments
AL2dns.i        Routine (~2D) MCNP5(X) input with neutron source
AL2dgs.i        Routine (~2D) MCNP5(X) input with gamma source
AL3dn.i         Detailed 3D MCNP5(X) input for neutron spectrum calc.
AL3dg.i         Detailed 3D MCNP5(X) input with DT source -gamma calc.
DT_MCNP5.TXT    patch for MCNP5 to calculate D-T neutron source
source.F        source.F for MCNPX to calculate D-T neutron source
srcdx.F         srcdx.F for MCNPX to calculate D-T neutron source
D-T.pdf         Document describing D-T source routine for MCNPX/MCNP5
mcnp4b.inp      1D MCNP-4B input model for (n,gamma) (OBSOLETE)
mcnp4b_g.inp    1D MCNP-4B input for gamma calculations (OBSOLETE)
okal-f1.gif     Fig. 1: Aluminium sphere geometry
okal-f2.gif     Fig. 2: Experimental setup
okal-f1.tif     Fig. 1: Al sphere geometry (high quality TIF)
okal-f2.tif     Fig. 2: Experimental setup (high quality TIF)
okal-f3.gif     Fig. 3: Neutron Leakage Spectra (low energy) -1D model
okal-f4.gif     Fig. 4: Neutron Leakage Spectra (high energy)-1D model
okal-f5.gif     Fig. 5: Neutron Leakage Spectra (full range) -1D model
okal-f6.gif     Fig. 6: Gamma Leakage Spectra (low energy) -1D model
okal-f7.gif     Fig. 7: Gamma Leakage Spectra (high energy) -1D model
j94-014.pdf     Reference 5
j98-024.pdf     Reference 7
o-c83-01.pdf    Reference 8
ane-10.pdf      Reference 11

NEA-1553/78
okw-abs.htm     Abstract
okmn-exp.htm    Description of experiment
Figure 1        Experimental arrangement of the OKTAVIAN Facility
Figure 2        Type I vessel
Mn2dn.i         Routine (~2D) MCNP5(X) input with neutron source
Mn3dn.i         Detailed 3D MCNP5(X) input for neutron spectrum analysis
j94-014.pdf     Reference
j98-024.pdf     Reference
Oktavian.pdf    Document describing quality assessment of OKTAVIAN experiments
top ]
17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: benchmarks, experiment, fusion reactions, neutron.