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FNG-ITER DOSE RATE EXPERIMENT
1. Name of Experiment:
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FNG-ITER Dose Rate Experiment
(2000-2001)
2. Purpose and Phenomena Tested:
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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.
3. 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.
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).
4. Measurement System:
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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.
5. 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.
6. Special Features:
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None
7. Author/Organizer:
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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
E-mail: batiston at efr406.frascati.enea.it
Reviewer of compiled data:
I. Kodeli
OECD/NEA, 12 bd des Iles, 92130 Issy les Moulineaux, France
e-mail: ivo.kodeli at oecd.org
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).
8. Availability:
------------
Unrestricted
9. References:
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[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
[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;
[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))
[7] Briesmeister, J. F. (Ed.), MCNP - A general Monte Carlo n-particle
transport code, version 4C, Report LA12625, Los Alamos, September
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.
10. Data and Format:
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FILE NAME bytes Content
---- ------------ ------ -------
1 fngdos-a.htm 16,272 This information file
2 fngdos-e.htm 75,464 Description of Experiment
3 mcnp_n.inp 49,281 3-D model for MCNP-4C calculations of neutron
flux (irradiation model)
4 mcnp_g.inp 67,976 3-D model for MCNP-4C calculations of dose rate
(decay gamma transport, shut down model)
5 fisp_620.inp 1,014 Input for FISPACT run (e.g. for cell 620)
6 fisp_620.flx 1,827 Neutron flux, calculated in mcnp_n.inp, for
FISPACT run (e.g. for cell 620)
7 fisp_620.out 843,563 Output of FISPACT run (e.g. for cell 620)
containing decay g-ray spectra to be input in
mcnp_g.inp
8 source.for 45,178 FORTRAN subroutine for MCNP source description
9 fig1.gif 5,242 Fig. 1: Angular dependence of the source
10 fig2.gif 9.505 Fig. 2: Energy/angular dependence of the source
11 fig3.gif 16,922 Fig. 3: Geometry of the experimental mock-up
12 fig4.gif 10,592 Fig. 4: Nickel activation foil positions in the
cavity
13 fig5.gif 20,177 Fig. 5: Neutron irradiaton time profile
14 fig6.gif 15,487 Fig. 6: Percent contributions of most important
radioisotopes to the total contact dose rate,
calculated by FISPACT for cell 620 of mcnp_n.inp.
15 fig7.gif 13,405 Fig. 7: Measured dose rate in the cavity centre
as a function of cooling time.
16 fig8.gif 9,633 Fig. 8: Geometry of the TiT target
17 fng-dose.pdf 2,110,390 Ref. 1
18 nd2001.pdf 330,452 Ref. 2
19 eff-726.pdf 161,239 Ref. 3
20 eff-614.pdf 585,759 Ref. 6
File fngdos-e.htm contains the following tables:
Tab. 1: Angular dependence of the source
Tab. 2: Angular/energy dependence of the source energy distribution
Tab. 3: Geometrical arrangement of the bulk shielding assembly
Tab. 4: Chemical composition of stainless steel SS316
Tab. 5: Irradiation history
Tab. 6: Measured (E) dose rates inside the cavity (G-M)
Tab. 7: Measured (E) dose rates in the cavity (TLD)
Tab. 8: Measured (E) Ni-58(n,p)Co-58 reaction rates
Tab. 9: Measured (E) Ni-58(n,2n)Co-58 reaction rates
Tab.10: Calculated dose rate (C) - Comparison between calculated
and measured values (C/E ratios)
Tab.11: Comparison between calculated and measured Ni-58(n,p)
reaction rates (C/E ratios).
Tab.12: Comparison between calculated and measured Ni-58(n,2n)
reaction rates (C/E ratios).
The figures are given in gif format.
SINBAD Benchmark Generation Date: 4/2003
SINBAD Benchmark Last Update: 4/2003
