Sensitivity and uncertainty analysis performed on a 14-MeV neutron streaming experiment (original) (raw)
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Neutron streaming experiment at FNG: results and analysis
Fusion Engineering and Design, 2000
The experimental validation of the available codes and nuclear data to calculate correctly the effects of streaming paths in the shield of a fusion reactor was provided by a streaming experiment carried out at the 14 MeV Frascati Neutron Generator (FNG). The experiment consisted in the irradiation of a mock-up of the shielding blanket (stainless steel and water equivalent material, 1 m thick) and of the toroidal field coil. A void channel with high aspect ratio and a cavity at the end of the channel were realised in the bulk shield. Some nuclear quantities were measured in the cavity at the end of the channel and behind it up to the simulated superconducting magnet. The neutron flux was measured using the activation foil technique while the nuclear heating was measured using thermoluminescent detectors. The experimental data were compared with the results of calculations performed by the Monte Carlo code MCNP-4B, using the Fusion Evaluated Nuclear Data Library (FENDL, version 1.0, 2.0) and the European Fusion File (EFF, versions 3.0, 3.1). The comparison shows that all the used libraries reproduce the neutron flux generally within 920% total uncertainty both in the void penetration and in the bulk shield behind it. The trend to underestimate the high energy flux (E\ 8 MeV) with increasing penetration depth is found again as in previous bulk shield experiment, showing that the relevant transport cross sections at around 14 MeV need further investigation. The comparison between experimental and calculated nuclear heating shows an agreement within 9 30% for all tested nuclear data libraries and in all experimental positions. The comparison with the case of bulk shield without penetration is also discussed.
Fusion Engineering and Design, 2004
A sensitivity analysis has been performed for a 14 MeV neutron benchmark on an iron assembly, typical for a fusion neutronic integral experiment. Probabilistic and deterministic computational methods have been used in the sensitivity calculations with the main objective to check and validate the novel Monte Carlo technique for calculating point detector sensitivities. Good agreement has been achieved between the Monte Carlo and the deterministic approaches for the individual calculated sensitivity profiles, the uncertainties and the neutron flux spectra. It is thus concluded that the Monte Carlo technique for calculating point detector sensitivities and related uncertainties as being implemented in MCSEN, a local version of the MCNP4A code with the capability to calculate point detector sensitivities, is well qualified for sensitivity and uncertainty analyses of integral experiments.
Nuclear Data Sensitivity/Uncertainty Pre-Analysis of FNG WCLL Fusion Benchmark
EPJ Web of Conferences
To assure tritium self-sufficiency in future fusion reactors such as DEMO the accuracy of TRP calculations has to be demonstrated within the design uncertainties. A new neutronics experiment representing a mock-up of the Water Cooled Lithium Lead (WCLL) Test Blanket Module (TBM) is under preparation at the Frascati neutron generator (FNG) with the objective to provide an experimental validation of accuracy of nuclear data and neutron transport codes for the tritium production rate (TPR) calculations. The mock-up will consist of LiPb bricks, EUROFER plates and Perspex substituting water. The mock-up will be irradiated by 14 MeV neutrons at the FNG facility, and the TPR and detector reaction rates will be measured using Li2CO3 pellets and activation foils placed at different positions up to about 55 cm inside the mock-up. Computational pre-analyses for the design of the WCLL neutronics experiment using the SUSD3D sensitivity/uncertainty (S/U) code system is described and compared with...
Characterization of Fast Neutron Transmission Through an Iron Shield
Energija, 2023
In this paper we give an analysis of the neutron transmission through an iron sphere using Monte Carlo and transport theory methods based on ENDF/B-VII.1 general purpose library. The motivation for this investigation comes from a well-known deficiency in the iron inelastic data from the older library evaluation (ENDF/B-V), giving a concern for a fast neutron flux underestimation within the reactor pressure vessels. In order to benchmark the next-generation ENDF/B-VI iron data, the U.S. Nuclear Regulatory Commission and the former Czechoslovakian National Research Institute have jointly preformed several experiments in 1990s, addressing neutron leakage spectra obtained for a 252 Cf fission source in a centre of an iron sphere. It was shown that the ENDF/B-VI iron cross section, containing several improvements over previous evaluations, will not entirely resolve the neutron spectrum discrepancies observed at high neutron energies. Since safety analyses of reactor pressure vessel embrittlement are often based on neutron transport calculations using specific multigroup cross section libraries, simulation of this benchmark was performed using a hybrid shielding methodology of ADVANTG3.0.3 and MCNP6.1.1b codes. Comparison of calculated and referenced dosimeter activation rates are presented for several »standard« nuclear reactions, often used in reactor pressure vessel dosimetry. For that purpose, the new IRDFF-II special library from the IAEA Nuclear Data Services was used as a reference source of dosimetry cross sections. The MCNP6.1.1b code was used for calculation of reaction rates, which were also compared with previous IRDFF-1.05 special library and general purpose ENDF/B-VII.1 library.
Fusion Engineering and Design, 2000
The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium-lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens' irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties. The analysis has been carried out using the LSL-M2 code, which optimizes the neutron spectrum by means of a least-squares technique taking into account the variance and covariance files. In the second part of the activity, the possibility of extending to IFMIF the use of existing on-line in-core neutron/gamma monitors (to be located at several positions inside the IFMIF test cell for beam control, safety and diagnostic purposes) has been studied. A feasibility analysis of the modifications required to adapt sub-miniature fission chambers (recently developed by CEA-Cadarache) to the high flux test module of the test cell has been carried out. The verification of this application pertinence and a gross definition of the in-core detector characteristics are described. The option of using self-powered neutron detectors (SPNDs) is also discussed.
Detector sensitivity and calibration for yield measurements in fusion neutron sources
Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2019
In this study, numerical simulations have been carried out to evaluate the effect of moderator parameters (material/dimensions) for standard radio-isotopic/fission (Pu-Be and 252 Cf) and fusion sources (D-D). These sources are used for calibrating thermal neutron detectors, which are often employed for determining absolute neutron yield in fusion-based sources. Simulations have been carried out using the Monte Carlo based ''FLUKA'' code for typical moderator materials and 3 He/BF 3 proportional counter based thermal neutron detectors. Results have been validated against experimentally measured values for an available radio-isotopic source. A notable difference in the detector efficiency was observed due to the source energy term. The detection efficiency ratio for Pu-Be:D-D: 252 Cf sources with the same moderator thickness (optimized for 252 Cf) was found to be 1 : 1.16±0.02 : 1.40±0.03. For these three types of sources, optimum acrylic (Perspex) and polypropylene moderator thicknesses have been observed to be 100,100,80 mm and 80,80,60 mm respectively with the corresponding efficiency ratio of 1 : 1.15±0.02 : 1.34±0.008.
Fusion Technology, 1988
Several integral experiments on tritium breeding were jointly performed at the Fusion Neutronics Source (FNS) facility at the Japan Atomic Energy Research Institute (JAERI), in connection with the U. S./JAERI Collaborative Program on Fusion Breeder Neutronics. Tritium production rates from 6 Li (T 6) and 7 Li (T 7) were measured at several locations in an Li 2 0 assembly (D = 60 cm, L = 60 cm) embedded in the concrete wall of a 5-X 5x 4.5-m room (reference experiment). JAERI has also performed independent benchmark experiments with the Li 2 0 assembly located in a large room of negligible room-return neutrons. In the reference experiment, large discrepancies in T 6 were found at the front locations in the Li 2 0 assembly. At middle locations, the calculated-to-experimental (C/E) values for T 6 are-1.2 (U.S.) and-1.1 (JAERI). The C/E values for T 7 are ~1.18 (U.S.) and 1.05 (JAERI). To assess the contribution to the uncertainty in predicting T 6 and T 7 that results from the current uncertainties in the nuclear data base, an extensive two-dimensional cross-section sensitivity/uncertainty analysis was performed. For that purpose, the FORSS module, and the VIP and DOT 4.3 codes were used along with the PUFF-2 covariance code. Two systems were considered for the analysis: the benchmark system and the reference system. The models used simulate the geometrical details and source conditions for the experiments. After coupling the sensitivity profiles with the cross-section uncertainty information (ENDF/B-V, file 33), it was found that the standard deviations in T 6 are 2.0 to 3.5%. In the reference system, the uncertainties in T 6 at front locations due to data uncertainties were found to be very small (-0.3%). The large discrepancies at these locations between the calculation and measurements were attributed to inaccuracy in modeling and predicting the room-return component of incident neutrons. The uncertainties in T 7 due to the uncertainties in nuclear data were found to be 3 to 6%, with the largest values at back locations. The discrepancies with experimental values were attributed to the inaccuracy in the ? Li(n,n'cc)t cross section, which requires further evaluation.
The bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)
Fusion Engineering and Design, 1995
In the design of next-step fusion devices such as NET/ITER the nuclear performance of shielding blankets is of key importance in terms of nuclear heating of superconducting magnets and radiation damage. In the framework of the European Fusion Technology Program, ENEA Frascati and CEA Cadarache in collaboration performed a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG), aimed at obtaining accurate experimental data for improving the nuclear database and methods used in shielding designs. The experiment consisted of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron reaction rates at various depths inside the block have been measured using fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both Sn and Monte Carlo transport codes and the cross-section library EFF.1 (European Fusion File). 0920-3796/95/$09.50 © 1995 Elsevier Science S.A. All rights reserved SSDI 0920-3796(94) 00093-X
Nuclear Fusion, 2017
A new DT campaign (DTE2) is planned at JET in 2020 to minimize the risks of ITER operations. In view of DT operations, a calibration of the JET neutron monitors at 14 MeV neutron energy has been performed using a well calibrated 14-MeV neutron generator (NG) deployed, together with its power supply and control unit, inside the vacuum vessel by the JET remote handling system. The neutron generator was equipped with two calibrated diamond detectors, which continuously monitored its neutron emission rate during the calibration, and activation foils which provided the time integrated yield. Cables embedded in the remote handling boom were used to power the neutron generator, the active detectors and pre-amplifier, and to transport the detectors' signal. The monitoring activation foils were retrieved at the end of each day for decay -ray counting, and replaced by fresh ones. About 76 hours of irradiation, in 9 days, were needed with the neutron generator in 73 different poloidal and toroidal positions in order to calibrate the two neutron yield measuring systems available at JET, the 235 U fission chambers (KN1) and the inner activation system (KN2). The NG neutron emission rates provided by the monitoring detectors were in agreement within 3-4%. Neutronics calculations have been performed using MCNP code and a detailed model of JET to derive the response of the JET neutron detectors to DT plasma neutrons starting from the response to the NG neutrons, and taking into account the anisotropy of the neutron generator and all the calibration circumstances. These calculations have made use of a very detailed and validated geometrical description of the neutron generator and of the neutron source routine producing neutron energy-angle distribution for the neutron emitted by the neutron generator. The KN1 calibration factor for a DT plasma has been determined with ±4.2% experimental uncertainty (1). It has been found that the difference between KN1 response to DD neutrons and that to DT neutrons is within the uncertainties in the derived responses. KN2 has been calibrated using the 93 Nb(n,2n) 92m Nb and 27 Al(n,a) 24 Na activation reactions (energy thresholds 10 MeV and 5 MeV respectively). The total uncertainty on the calibration factors is ±6% for 93 Nb(n,2n) 92m Nb and ±8% 27 Al(n,a) 24 Na (1). The calibration factors of the two independent systems KN1 and KN2 will be validated during DT operations. The experience gained and the lessons learnt are presented and discussed in particular with regard to the 14 MeV neutron calibrations in ITER.
CEA develops and makes use of miniature fission chambers (MFCs, with radius down to 1.5 mm) for reactor physics conducted in experimental reactors such as EOLE and MINERVE zero power reactors (CEA Cadarache). When measuring fission rate, it is known that the neutron spectrum in the irradiation channel can be modified by the detector and the detector fixture. So the result of the measurement does not give a direct access to the desired quantity (fission rate, neutron flux,etc.) To overcome this problem, it is possible to make use of Monte Carlo calculations based on a detailed modeling of the detector. It could then be included in the 3D reactor model but this leads to large and time consuming calculations. In this case, measurement results can be combined directly with calculated values to produce the desired quantity. Another possibility is to calculate correction factors to apply to the biased measurement, i.e. to perform two-step calculations. Those factors depend on the detector...