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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
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Examination of Mars2020 Shock-Layer Conditions Via Infrared Emission Spectroscopy of CO2

A new test series was performed in the Electric Arc Shock Tube (EAST) facility, with the aim of reproducing flight conditions encountered during the Mars2020 mission entry into Mars atmosphere. For this test series, the EAST facility was instrumented with two spectrometers for Optical Emission Spectroscopy (OES) measurements and three mid-infrared lasers for Tunable Diode Laser Absorption Spectroscopy (TDLAS). This study focuses on the spectrally and spatially resolved radiance measured with OES. Comparison with CEA/NEQAIR and DPLR/NEQAIR simulations were made. For velocity above 2.7 km/s, measurements are found to lie between the different tested kinetic models, within ± 10%. Below 2.7 km/s, the chemistry appears to be frozen and measured radiance profiles are typically 10% above CEA predictions. This difference is explained by shock deceleration effects. Finally, comparison between OES and TDLAS data are made and show a good agreement on the measured temperature and CO2 number density profiles using the two diagnostics, typically within 10 and 5%, respectively.

Augustin Tibère-Inglesse↗
DOE OSTI2011

JINA Workshop Nuclear Physics in Hot Dense Dynamic Plasmas

Measuring NEET and NEEC is relevant for probing stellar cross-sections and testing atomic models in hot plasmas. Using NEEC and NEET we can excite nuclear levels in laboratory plasmas: (1) NIF: Measure effect of excited nuclear levels on (n,{gamma}) cross-sections, 60% and never been measured; (2) Omega, Test cross-sections for creating these excited levels via NEEC and NEET. Will allow us to test models that estimate resonance overlap of atomic states with the nucleus: (1) Average Atom model (AA) (CEA&LLNL), single average wave-function potential; (2) Super Transition Array (STA) model (LLNL), More realistic individual configuration potentials NEET experimental data is scarce and not in a plasma environment, NEEC has not yet been observed.

42 ENGINEERING↗
DOE OSTI2011

Summary of the First Neutron Image Data Collected at the National Ignition Facility

A summary of data and results from the first neutron images produced by the National Ignition Facility (NIF), Lawrence Livermore National Laboratory, Livermore, CA, USA are presented. An overview of the neutron imaging technique is presented, as well as a synopsis of the data collected and measurements made to date. Data form directly driven, DT filled microballoons, as well as, indirectly driven, cryogenically layered ignition experiments are presented. The data presented show that the primary cores from directly driven implosions are approximately twice as large, 64 +/- 3 um, as indirect cores (25 +/- 4 and 29 +/- 4 um and more asymmetric, P2/P0 = 47% vs. -14% and -7%. Further, comparison with the size and shape of X-ray image data from on the same implosions show good agreement, indicating X-ray emission is dominated by the hot regions of the implosion. This work was performed for the U.S. Department of Energy, National Nuclear Security Administration and by the National Ignition Campaign partners; Lawrence Livermore National Laboratory (LLNL), University of Rochester -Laboratory for Laser Energetics (LLE), General Atomics(GA), Los Alamos National Laboratory (LANL), Sandia National Laboratory (SNL). Other contributors include Lawrence Berkeley National Laboratory (LBNL), Massachusetts Institute of Technology (MIT), Atomic Weapons Establishment (AWE), England, and Commissariat `a l’ ´ Energie Atomique (CEA), France.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Benchmarking Of Improved DPAC Transient Deflagration Analysis Code

The transient deflagration code DPAC (Deflagration Pressure Analysis Code) has been upgraded for use in modeling hydrogen deflagration transients. The upgraded code is benchmarked using data from vented hydrogen deflagration tests conducted at the HYDRO-SC Test Facility at the University of Pisa. DPAC originally was written to calculate peak deflagration pressures for deflagrations in radioactive waste storage tanks and process facilities at the Savannah River Site. Upgrades include the addition of a laminar flame speed correlation for hydrogen deflagrations and a mechanistic model for turbulent flame propagation, incorporation of inertial effects during venting, and inclusion of the effect of water vapor condensation on vessel walls. In addition, DPAC has been coupled with CEA, a NASA combustion chemistry code. The deflagration tests are modeled as end-to-end deflagrations. The improved DPAC code successfully predicts both the peak pressures during the deflagration tests and the times at which the pressure peaks.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Development of Yield and Tensile Strength Design Curves for Alloy 617

The U.S. Department of Energy Very High Temperature Reactor Program is acquiring data in preparation for developing an Alloy 617 Code Case for inclusion in the nuclear section of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. A draft code case was previously developed, but effort was suspended before acceptance by ASME. As part of the draft code case effort, a database was compiled of yield and tensile strength data from tests performed in air. Yield strength and tensile strength at temperature are used to set time independent allowable stress for construction materials in B&PV Code, Section III, Subsection NH. The yield and tensile strength data used for the draft code case has been augmented with additional data generated by Idaho National Laboratory and Oak Ridge National Laboratory in the U.S. and CEA in France. The standard ASME Section II procedure for generating yield and tensile strength at temperature is presented, along with alternate methods that accommodate the change in temperature trends seen at high temperatures, resulting in a more consistent design margin over the temperature range of interest.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Prototype HB650 Transportation Validation For The PIP-II Project

The PIP-II Project at Fermilab is centered around a superconducting 800 MeV proton linac to upgrade and modernize the Fermilab accelerator complex, allowing increased beam current to intensity frontier experiments such as LBNF-DUNE. PIP-II includes strong international collaborations, including the delivery of 12 cryomodules from European labs to FNAL (3 from STFC-UKRI in the UK and 9 from CEA in France). The transatlantic shipment of these completed modules is identifed as a serious risk for the project. To mitigate this risk, a rigorous and systematic process has been developed to design and validate a transport system, including specifcation, procedures, logistics, and realistic testing. This paper will detail the engineering process used to manage this effort across the collaboration and the results of the first major validation testing of the integrated shipping system prior to use with a cryomodule.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2011

Development of high power CW 3.7 GHz klystrons for fusion experiments on Tore Supra

In the frame of the CIMES project, a collaborative effort between Association Euratom-CEA and Thales Electron Devices (TED) has led to the development of a high power CW klystron TH 2103 C, working at 3.7 GHz, for plasma heating and current drive for the Tokamak Tore Supra. A prototype has been manufactured and thoroughly tested on water load in December 2007 to verify that all the parameters met the specifications. The paper will present in detail the process and results of the test of the klystrons.

43 PARTICLE ACCELERATORS↗
DOE OSTI2012

Experimental validation of the DARWIN2.3 package for fuel cycle applications

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Burst wait time simulation of CALIBAN reactor at delayed super-critical state

In the past, the super prompt critical wait time probability distribution was measured on CALIBAN fast burst reactor [4]. Afterwards, these experiments were simulated with a very good agreement by solving the non-extinction probability equation [5]. Recently, the burst wait time probability distribution has been measured at CEA-Valduc on CALIBAN at different delayed super-critical states [6]. However, in the delayed super-critical case the non-extinction probability does not give access to the wait time distribution. In this case it is necessary to compute the time dependent evolution of the full neutron count number probability distribution. In this paper we present the point model deterministic method used to calculate the probability distribution of the wait time before a prescribed count level taking into account prompt neutrons and delayed neutron precursors. This method is based on the solution of the time dependent adjoint Kolmogorov master equations for the number of detections using the generating function methodology [8,9,10] and inverse discrete Fourier transforms. The obtained results are then compared to the measurements and Monte-Carlo calculations based on the algorithm presented in [7]. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Nuclear data uncertainty analysis for the generation IV gas-cooled fast reactor

For the European 2400 MW Gas-cooled Fast Reactor (GoFastR), this paper summarizes a priori uncertainties, i.e. without any integral experiment assessment, of the main neutronic parameters which were obtained on the basis of the deterministic code system ERANOS (Edition 2.2-N). JEFF-3.1 cross-sections were used in conjunction with the newest ENDF/B-VII.0 based covariance library (COMMARA-2.0) resulting from a recent cooperation of the Brookhaven and Los Alamos National Laboratories within the Advanced Fuel Cycle Initiative. The basis for the analysis is the original GoFastR concept with carbide fuel pins and silicon-carbide ceramic cladding, which was developed and proposed in the first quarter of 2009 by the 'French alternative energies and Atomic Energy Commission', CEA. The main conclusions from the current study are that nuclear data uncertainties of neutronic parameters may still be too large for this Generation IV reactor, especially concerning the multiplication factor, despite the fact that the new covariance library is quite complete; These uncertainties, in relative terms, do not show the a priori expected increase with bum-up as a result of the minor actinide and fission product build-up. Indeed, they are found almost independent of the fuel depletion, since the uncertainty associated with {sup 238}U inelastic scattering results largely dominating. This finding clearly supports the activities of Subgroup 33 of the Working Party on International Nuclear Data Evaluation Cooperation (WPEC), i.e. Methods and issues for the combined use of integral experiments and covariance data, attempting to reduce the present unbiased uncertainties on nuclear data through adjustments based on available experimental data. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Reactivity worth measurements on fast burst reactor Caliban - description and interpretation of integral experiments for the validation of nuclear data

Reactivity perturbation experiments using various materials are being performed on the HEU fast core CALIBAN, an experimental device operated by the CEA VALDUC Criticality and Neutron Transport Research Laboratory. These experiments provide valuable information to contribute to the validation of nuclear data for the materials used in such measurements. This paper presents the results obtained in a first series of measurements performed with Au-197 samples. Experiments which have been conducted in order to improve the characterization of the core are also described and discussed. The experimental results have been compared to numerical calculation using both deterministic and Monte Carlo neutron transport codes with a simplified model of the reactor. This early work led to a methodology which will be applied to the future experiments which will concern other materials of interest. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

An example of neutronic penalizations in reactivity transient analysis using 3D coupled chain HEMERA

HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), is a fully coupled 3D computational chain developed jointly by IRSN and CEA. It is composed of CRONOS2 (core neutronics, cross sections library from APOLLO2), FLICA4 (core thermal-hydraulics) and the system code CATHARE. Multi-level and multi-dimensional models are developed to account for neutronics, core thermal-hydraulics, fuel thermal analysis and system thermal-hydraulics, dedicated to best-estimate, conservative simulations and sensitivity analysis. In IRSN, the HEMERA chain is widely used to study several types of reactivity accidents and for sensitivity studies. Just as an example of the HEMERA possibilities, we present here two types of neutronic penalizations and their impact on a power transient due to a REA (Rod Ejection Accident): in the first one, we studied a bum-up distribution modification and in the second one, a delayed-neutron fraction modification. Both modifications are applied to the whole core or localized in a few assemblies. Results show that it is possible to use global or local changes but 1) in case of bum-up modification, the total core power can increase when assembly peak power decrease so, care has to be taken if the goal is to maximize a local power peak and 2) for delayed-neutron fraction, a local modification can have the same effect as the one on the whole core, provided that it is large enough. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

An improved energy-collapsing method for core-reflector modelization in SFR core calculations using the PARIS platform

In the framework of the ASTRID project, sodium cooled fast reactor studies are conducted at CEA in compliance with GEN IV reactors criteria, particularly for safety requirements. An improved safety requires better calculation tools to obtain accurate reactivity effects (especially sodium void effect) and power map distributions. The current calculation route lies on the JEFF3.1.1 library and the classical two-step approach performed with the ECCO module of the ERANOS code system at the assembly level and the Sn SNATCH solver - implemented within the PARIS platform - at the core level. 33-group cross sections used by SNATCH are collapsed from 1968-group self-shielded cross-section with a specific flux-current weighting. Recent studies have shown that this collapsing is non-conservative when dealing with core-reflector interface and can lead to reactivity discrepancies larger than 500 pcm in the case of a steel reflector. Such a discrepancy is due to the flux anisotropy at the interface, which is not taken into account when cross sections are obtained from separate fuel and reflector assembly calculations. A new approach is proposed in this paper. It consists in separating the self-shielding and the flux calculations. The first one is still performed with ECCO on separate patterns. The second one is done with SNATCH on a 1D traverse, representative of the core-reflector interface. An improved collapsing method using angular flux moments is then carried out to collapse the cross sections onto the 33-group structure. In the case of a simplified ZONA2B 2D homogeneous benchmark, results in terms of k{sub eff} and power map are strongly improved for a small increase of the computing time. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

SILER: Seismic-Initiated events risk mitigation in Lead-cooled Reactors

SILER is a Collaborative Project, partially funded by the European Commission, aimed at studying the risk associated to seismic initiated events in Generation IV Heavy Liquid Metal reactors and developing adequate protection measures. The attention is focused on the evaluation of the effects of earthquakes (with particular regards to beyond design seismic events) and to the identification of mitigation strategies, acting both on structures and components design (as well as on the development of seismic isolation devices) which can also have positive effects on economics, leading to an high level of plant design standardization. Attention is also devoted to the identification of plant layout solutions able to avoid risks of radioactive release from both the core and other structures (i.e. the spent fuel storage pools). Specific effort is paid to the development of guidelines and design recommendations for addressing the seismic issue in next generation reactor systems. In addition, consideration will be devoted to transfer the knowledge developed in the project to Generation III advanced systems, in line with the objective of the SNE-TP SRA to support present and future Light Water Reactors and their further development, for which safety issues are key aspects to be addressed. Note, in this respect, that the benefits of base isolation in terms of response to design seismic actions are already widely recognized for Generation III LWRs, along with the possibility of a significant standardization of structural and equipment design. SILER activities started on October 1 st 2011 and are carried out by 18 partners: ENEA (Italy, Coordinator), AREVA NP SAS (France), SCK-CEN (Belgium), FIP Industriale (Italy), MAURER SOHENE (Germany), EC-JRC (Ispra (Italy)), SINTEC (Italy), KTH (Sweden), BOA-BKT (Germany), IDOM (Spain), ANSALDO (Italy), IPUL (Latvia), NUMERIA (Italy), VCE (Austria), SRS (Italy), CEA (France), EA (Spain), NUVIA (France). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Interpretation of 3D void measurements with Tripoli4.6/JEFF3.1.1 Monte Carlo code

The present work details the first analysis of the 3D void phase conducted during the EPICURE/UM17x17/7% mixed UOX/MOX configuration. This configuration is composed of a homogeneous central 17x17 MOX-7% assembly, surrounded by portions of 17x17 1102 assemblies with guide-tubes. The void bubble is modelled by a small waterproof 5x5 fuel pin parallelepiped box of 11 cm height, placed in the centre of the MOX assembly. This bubble, initially placed at the core mid-plane, is then moved in different axial positions to study the evolution in the core of the axial perturbation. Then, to simulate the growing of this bubble in order to understand the effects of increased void fraction along the fuel pin, 3 and 5 bubbles have been stacked axially, from the core mid-plane. The C/E comparison obtained with the Monte Carlo code Tripoli4 for both radial and axial fission rate distributions, and in particular the reproduction of the very important flux gradients at the void/water interfaces, changing as the bubble is displaced along the z-axis are very satisfactory. It demonstrates both the capability of the code and its library to reproduce this kind of situation, as the very good quality of the experimental results, confirming the UM-17x17 as an excellent experimental benchmark for 3D code validation. This work has been performed within the frame of the V and V program for the future APOLL03 deterministic code of CEA starting in 2012, and its V and V benchmarking database. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity of the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Analysis of Phenix natural convection test with the TRACE code

Experimental data from the Natural Convection (NC) test performed in the Phenix reactor prior to its final shutdown have been used to further validate the single-phase sodium flow modeling in TRACE. The experimental data for the benchmark have been shared by the CEA in the frame of a Coordinated Research Project (CRP), initiated by the IAEA Technical Working Group on Fast Reactors (TWG-FR). This paper presents a complete TRACE model of the Phenix primary circuit developed for the analysis. Steady-state calculations at nominal (350 MWth) and reduced (120 MWth) power are compared to the experimental data for the validation of the model. We presents results from the 'blind' comparison, i.e. the comparison of the test results with those computed prior to the communication of the experimental data, so-called 'pre-test' results. 'Post-test' results, calculated from a model improved on the basis of the discrepancies identified from the blind comparison, are also presented. The analysis highlights the need to accurately simulate the reactor structures, since these define the thermal inertia of the system during the first phase of the transient. Furthermore, it shows the limitations of computed 1D-results when applied to the simulation of highly-stratified temperature fields. Nevertheless, the simulated reactor behavior and temperatures are found to match very well with the experimental data after the first two hours and, in general, the TRACE blind predictions may be considered as having been quite satisfactory. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

CATHARE thermal-hydraulic system code for HLM preliminary validation in natural convection tests

The innovative nuclear systems cooled by Heavy Liquid Metal (HLM) are the subject of an ongoing interest both in Europe and outside, evidenced by a number of projects in progress. In the frame of the European Framework Programmes have been evidenced the need to adopt a thermalhydraulic system code capable to treat lead and Lead-Bismuth Eutectic (LBE) systems, with a particular interest in developing a 'European' code. Considering this scenario, within a specific collaboration between ENEA and CEA, the CATHARE French system code has been modified extending its capabilities to simulate HLM systems. In the present paper, the state of the validation process of CATHARE-HLM is discussed. The activity aims to assess the capabilities and limitations of the code to simulate the behavior of integral facilities, in particular in natural circulation conditions. The experimental data come from NACIE LBE-cooled facility sited at the ENEA Brasimone laboratories. The results obtained show a good capability in reproducing the systems behavior, despite some uncertainties on the experimental measurements. Future improvements on the code are going to be planned within the collaboration ENEA/CEA. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗