SEACAB qualification with Frascati Neutron Generator residual dose measurements (original) (raw)
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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
Fusion Engineering and Design, 2016
With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D-T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. In general, relevant benchmark results were obtained. Both data libraries can thus be run with TRIPOLI-4 for the fusion neutronics study. This work also demonstrates that the "safety factors" concept is necessary in the nuclear analyses of ITER.
EPJ Web of Conferences, 2016
Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 S N code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.
Fusion Engineering and Design, 1998
The recently developed FENDL-1 database, both in multigroup form (FENDL/MG-1.0) and continuous energy form (FENDL/MC-1.0) has been tested through analyzing a fusion integral experiment performed at the FNS facility, Japan, on a large bulk shielding assembly made of multilayers of SS316 and water. The assembly is a replica that simulates ITER shielding blanket and is bombarded by a 14 MeV source placed at 30 cm from the cylindrical assembly and housed inside a SS316 cylindrical can. This activity is undertaken as part of co-operation with JAERI on executing the required neutronics R&D tasks for ITER shield design. The objectives are (a) benchmarking FENDL-1 data and identifying any flaws that may exist in this newly developed database, and (b) examining the range of discrepancy between the calculated nuclear parameters inside the assembly and the measured ones in terms of the ratio of the calculated-to-experimental (C/E) data. Both differential and integral experimental data were analyzed along the central axis of the 120 cm D× 140 cm L assembly. The analyses with the multigroup data, MG also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo (MC) code was used with the FENDL/MC-1 data. The largest range of discrepancy between calculated and measured responses (reaction rates, neutron spectra, gamma ray heating, etc.) was found to be 20 -30% even though in most cases this discrepancy falls within the experimental errors.
Validation and improvement of Fe, Cr, Ni nuclear data in bulk shield benchmark experiments
Fusion Engineering and Design, 1995
To check shielding calculation routes and nuclear data libraries against experiments, a database of international Fe benchmarks has been built up. Accurate reaction rate measurements inside a stainless steel block, performed in the recent FNG European experiment, were also used. Two-dimensional SN calculations and three-dimensional Monte Carlo calculations, based on the multigroup MATXS87/EFF1 library, were carried out. The calculation-experiment comparison demonstrated the overall validity of the European EFV-1 library for shielding design calculations; however, flux underestimation with penetration depth was pointed out. A sensitivity study of the 70 measured response functions to the Fe partial cross-sections was achieved. In order to interpret the calculation experiment discrepancies in the most efficient way, a "trend analysis" method was implemented, accounting for both microscopic and integral data uncertainties. The Fe elastic cross-section is overestimated by 5% in the 100 keV-2 MeV range, probably because of resonance self-shielding. A clear trend to reduce the inelastic scattering was found out, in agreement with the ENDF/B-6 evaluation.
Shutdown dose calculations for the IFMIF test facility and the high flux test module
Progress in Nuclear Science and Technology, 2014
This work presents neutronic analyses to support the design of the test facility of the International Fusion Materials Irradiation Facility (IFMIF) in the framework of the Broader Approach activities. In order to evaluate the necessary biological shielding thickness and the reliability of remote handling tools for activated components inside the Test Cell, a detailed activation analysis has been carried out to obtain a detailed residual dose map resulting from the entire High Flux Test Module and Test Cell by utilizing the McDeLicious-11 Monte Carlo code and the mesh-tally based rigorous 2-Step approach (R2Smesh) developed at KIT. The absorbed dose rate expected at the working place of remote handling tools inside the Test Cell is approximately 100 Gy/h. A wall thickness of 150 cm for heavy concrete can be used as the design basis for hot cells in the test facility.
SINBAD – Radiation shielding benchmark experiments
Annals of Nuclear Energy, 2021
The creation of an international shielding benchmark database was presented in 1988 at the International Reactor Shielding Conference (ICRS7) in Bournemouth, UK. M. Salvatores was among the authors of the proposal and had promoted and contributed to the project since the first initiatives, showed continued interest and encouraged the development of the database. He was Chairman of the Committee of Reactor Physics (NEACRP) for 2 years (1984-1985) and Chair of the Shielding benchmark group (1982-1988). In particular, he chaired two annual meetings in 1984 and 1985, called to initiate the collaborative programme on the analysis of shielding benchmarks for the validation of the JEF data files where the need to organize shielding benchmark was recognized and the presentation at ICRS7 defined the overall project. SINBAD officially started in the early 1990 0 s as a collaboration between the OECD/NEA Data Bank and RSICC with the goal to preserve the information on the performed radiation shielding benchmark experiments and make these available in a standardised form to the international community. One key point concerned the sensitivity and uncertainty analyses required to define their quality and figures of merit. The database comprises now 102 shielding benchmarks, divided into three categories, covering both low and intermediate energy particles applications: fission reactor shielding (48 benchmarks), fusion blanket neutronics (31), and accelerator shielding (23) benchmarks. The database is intended for different users, including nuclear data evaluators, computer code developers, experiment designers and university students. SINBAD is available from RSICC and from the NEA Data Bank. The database was extensively used within the scope of numerous national and international projects, such as PWR Pressure vessel surveillance, fusion programme (ITER reactor studies), different OECD Working Parties on Evaluation Cooperation (WPEC) Subgroups, nuclear data validation, IAEA nuclear data projects, etc. The history of the database and few examples of its use are illustrated, for cross-sections, response functions and covariance matrix validation.
EPJ Web of Conferences
The Working Party on International Nuclear Data Evaluation Co-operation Subgroup 47 (WPECSG47) entitled "Use of Shielding Integral Benchmark Archive and Database for Nuclear Data Validation" was organised from 2019 and 2022 with the objectives to promote more systematic and wider use of shielding benchmark experiments in nuclear data (ND) and transport code validation and development, to provide feedback on the Shielding Integral Benchmark Archive and Database (SINBAD), and to promote its further development in coordination with the Expert Group on Physics of Reactor Systems (EGPRS). Altogether 9 meetings, the large majority (8) held remotely, were organised during the past 3 years to discuss the experience on the use of SINBAD, evaluation of new benchmarks and improvements to be contributed to the database which was severely neglected and lacking maintenance over the past ⇠ 10+ years. Several proposals for new or updated benchmark evaluation were presented and discussed, ...
Comparison of Computational and Experimental Dose Rates in a Neutron Activation Analysis Facility
2020
The vast majority of radiation protection guidelines in nuclear facilities usually relate from one to a few sources of radiation in very controlled environments. Currently, there are 111 research reactors where neutron activation analysis (NAA) is a major research and teaching component. In particular, NAA can yield a wide variety of exposures due to different types of samples and neutron fluxes. Unlike any other type of radiation laboratories, an NAA facility can contain a large variety of radioactive isotopes as a result of activation products with varying degrees of half-lives and with different intensities of gamma-rays and beta particles. Using MCNP 6.2, a Monte Carlo code developed by Los Alamos National Laboratory (LANL) for radiation transport, dose rates were computed. The computational results were validated by irradiating several National Institute of Standards and Technology (NIST) standard reference materials. The samples were allowed to decay during their transfer from the reactor to the NAA laboratory. These computational doses were validated to the experimental doses. Using this information, a database will be developed for accurately predicting the expected doses to researchers working at research reactors and develop better radiation protection standards at NAA facilities.