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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Reactor physics teaching and research in the Swiss nuclear engineering master

Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Preliminary resolved resonance region evaluation of copper-63 from 0 to 300 keV

A new preliminary evaluation of Cu-63 was done in the energy region from 0 to 300 keV extending the resolved resonance region of the previous, ENDF/B-VII.0, evaluation three-fold. The new evaluation was based on three experimental transmission data sets; two measured at the Oak Ridge Electron Linear Accelerator (ORELA) and one from the Massachusetts Inst. of Technology Nuclear Reactor (MITR). A total of 275 new resonances were identified and a corresponding set of external resonances was approximated to mock up the external levels. The negative external levels (bound level) were modified to match the thermal cross section values. A preliminary benchmarking calculation was made using 11 ICSBEP benchmarks. This work is in support of the DOE Nuclear Criticality Safety Program. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Verification study of thorium cross section in MVP calculation of thorium based fuel core using experimental data

Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Thermal neutron capture cross section of gadolinium by pile-oscillation measurements in MINERVE

Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of {sup 157}Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of {sup nat}Gd which is (49360 {+-} 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1{sigma}, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 {+-} 500) b. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Uncertainty analysis of a SFR core with sodium plenum

The new concepts of Sodium-cooled Fast Reactors have to reach the Generation IV safety objectives. In this regard the Sodium Void Effect has to be minimized for the future projects of large-size SFR as well as the uncertainties on it. The Inst. of Radiological Protection and Nuclear Safety (IRSN) as technological support of French public authorities is in charge of safety assessment of operating and under construction reactors, as well as future projects. In order to state about the safety of new SFR designs the IRSN must be able to evaluate core parameters and their uncertainties. In this frame a sensitivity and uncertainty study has been performed to evaluate the impact of nuclear data uncertainty on sodium void effect, for the benchmark model of large SFR BN-800. The benchmark parameters (effective multiplication factor and sodium void effect) have been evaluated using two codes, the deterministic code ERANOS and the Monte Carlo code SCALE, while the S/U analysis has been performed only with SCALE. The results of the these studies point out the most relevant cross section uncertainties that affect the SVE and how efforts should be done in increasing the existing nuclear data accuracies. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Transition to CASMO-5M and SIMULATE-3K for stability analyses of the Swiss BWRs

For stability analyses of the Swiss operating Boiling-Water-Reactors (BWRs), the methodology employed and validated so far at the Paul Scherrer Inst. (PSI) was based on the RAMONA-3 code with a hybrid upstream static lattice/core analysis approach using CASMO-4 and PRESTO-2. More recently, steps were undertaken towards a new methodology based on the SIMULATE-3K (S3K) code for the dynamical analyses combined with the CMSYS system relying on the CASMO/SIMULATE-3 suite of codes and which was established at PSI to serve as framework for the development and validation of reference core models of all the Swiss reactors and operated cycles. This paper presents a first validation of the new methodology on the basis of a benchmark recently organised by a Swiss utility and including the participation of several international organisations with various codes/methods. Now in parallel, a transition from CASMO-4E (C4E) to CASMO-5M (C5M) as basis for the CMSYS core models was also recently initiated at PSI. Consequently, it was considered adequate to address the impact of this transition both for the steady-state core analyses as well as for the stability calculations and to achieve thereby, an integral approach for the validation of the new S3K methodology. Therefore, a comparative assessment of C4 versus C5M is also presented in this paper with particular emphasis on the void coefficients and their impact on the downstream stability analysis results. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Advanced fuel assembly characterization capabilities based on gamma tomography at the Halden boiling water reactor

Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Inst. for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala Univ.. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Experimental power density distribution benchmark in the TRIGA Mark II reactor

In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Analysis of pin removal experiments conducted in an SCWR-like test lattice

A comprehensive program of integral experiments, largely based on the measurement of reaction rate distributions, was carried out recently on an SCWR-like fuel lattice in the central test zone of the PROTEUS zero-power research reactor at the Paul Scherrer Inst. in Switzerland. The present paper reports on the analysis of a complementary set of measurements, in which the reactivity effects of removing individual pins from the unperturbed, heterogeneously moderated reference lattice were investigated. It has been found that the detailed Monte Carlo modeling of the whole reactor using MCNPX is able - as in the case of the reaction rate distributions - to reproduce the experimental results for the pin removal worths within the achievable statistical accuracy. A comparison of reduced-geometry calculations between MCNPX and the deterministic LWR assembly code CASMO-4E has revealed certain discrepancies. On the basis of a reactivity decomposition analysis of the differences between the codes, it has been suggested that these could be due to CASMO-4E deficiencies in calculating the effect, upon pin removal, of the extra moderation in the neighboring fuel pins. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

On the multidimensional modeling of fluid flow and heat transfer in SCWRS

The Supercritical Water Reactor (SCWR) has been proposed as one of the six Generation IV reactor design concepts under consideration. The key feature of the SCWR is that water at supercritical pressures is used as the reactor coolant. Although at such pressures, fluids do not undergo phase change as they are heated, the fluid properties experience dramatic variations throughout what is known as the pseudo-critical region. Highly nonuniform temperature and fluid property distributions are expected in the reactor core, which will have a significant impact on turbulence and heat transfer in future SCWRs. The goal of the present work has been to understand and predict the effects of these fluid property variations on turbulence and heat transfer throughout the reactor core. Spline-type property models have been formulated for water at supercritical pressures in order to include the dependence of properties on both temperature and pressure into a numerical solver. New models of turbulence and heat transfer for variable-property fluids have been developed and implemented into the NPHASE-CMFD software. The results for these models have been compared to experimental data from the Korea Atomic Energy Research Inst. (KAERI) for various heat transfer regimes. It is found that the Low-Reynolds {kappa}-{epsilon} model performs best at predicting the experimental data. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

On-line maintenance methodology development and its applications

With the increasing economic pressures being faced and the potential for shortening outage times under the conditions of deregulated electricity markets in the world, licensees are motivated to get an increasing amount of the on-line maintenance (OLM). The benefits of the OLM includes increased system and plant reliability, reduction of plant equipment and system material condition deficiencies that could adversely impact operations, and reduction of work scope during plant refueling outages. In Korea, allowance guidelines of risk assessment is specified in the safety regulation guidelines 16.7 and 16.8 of the Korea Inst. of Nuclear Safety (KINS), which is 'General guidelines of Risk-informed application for requesting permission of changes' and 'Requesting permission of changes of Risk-informed application for Technical Specification'. We select the emergency diesel generator (EDG) of the Ulchin unit 3 and 4 for risk assessment analysis by applying configuration changes. The EDG which has plant safety level IE belongs to on-site standby power (A, B train EDG) in electric distribution system. The EDG is important component because it should maintain standby status during plant is operating, therefore we select the EDG for target component of risk assessment analysis. The risk assessment is limited to CDF. The risk assessment is performed by using AIMS-PSA Release2. We evaluate CDF by applying the configuration changes with some assumptions. Evaluation of the full power operation and Low power/Shut down operation was performed. This study has been performed for introducing a methodology and performing risk assessment. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Innovative applications of technology for nuclear power plant productivity improvements

The nuclear power industry in several countries is concerned about the ability to maintain high plant performance levels due to aging and obsolescence, knowledge drain, fewer plant staff, and new requirements and commitments. Current plant operations are labor-intensive due to the vast number of operational and support activities required by commonly used technology in most plants. These concerns increase as plants extend their operating life. In addition, there is the goal to further improve performance while reducing human errors and increasingly focus on reducing operations and maintenance costs. New plants are expected to perform more productively than current plants. In order to achieve and increase high productivity, it is necessary to look at innovative applications of modern technologies and new concepts of operation. The Electric Power Research Inst. is exploring and demonstrating modern technologies that enable cost-effectively maintaining current performance levels and shifts to even higher performance levels, as well as provide tools for high performance in new plants. Several modern technologies being explored can provide multiple benefits for a wide range of applications. Examples of these technologies include simulation, visualization, automation, human cognitive engineering, and information and communications technologies. Some applications using modern technologies are described. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Hydrogen combustion in a flat semi-confined layer with respect to the Fukushima Daiichi accident

The hydrogen accumulation at the top of containment or reactor building may occur due to an interaction of molten corium and water followed by a severe accident of a nuclear reactor (TMI, Chernobyl, Fukushima Daiichi). The hydrogen, released from the reactor, accumulates usually as a stratified semi-confined layer of hydrogen-air mixture. A series of large scale experiments on hydrogen combustion and explosion in a semi-confined layer of uniform and non-uniform hydrogen-air mixtures in presence of obstructions or without them was performed at the Karlsruhe Inst. of Technology (KIT). Different flame propagation regimes from slow subsonic to relative fast sonic flames and then to the detonations were experimentally investigated in different geometries and then simulated with COMSD code with respect to evaluate amount of burnt hydrogen taken place during the Fukushima Daiichi Accident (FDA). The experiments were performed in a horizontal semi-confined layer with dimensions of 9x3x0.6 m with/without obstacles opened from below. The hydrogen concentration in the mixtures with air was varied in the range of 0-34 vol. % without or with a gradient of 0-60 vol. %H{sub 2}/m. Effects of hydrogen concentration gradient, thickness of the layer, geometry of the obstructions, average and maximum hydrogen concentration on flame propagation regimes were investigated with respect to evaluate the maximum pressure loads of internal structures. Blast wave strength and dynamics of propagation after explosion of the layer of hydrogen-air mixture was numerically simulated to reproduce the hydrogen explosion process during the Fukushima Daiichi Accident. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Development of 3D full-core ERANOS-2.2/MCNPX-2.7.0 models and neutronic analysis of the BFS-2 zero-power facility

The present paper is addressing the development and validation against experimental data of 3D full-core models of the BFS-2 zero-power fast-reactor using both the deterministic system code ERANOS-2.2 and the stochastic code MCNPX-2.7.0. The model configuration of BFS considered for analysis is the BFS-62-3A benchmark. To extend the - deterministic/stochastic - code-to-code comparison, neutronic parameters, i.e. reactivity, neutron spectrum and reaction rates, were also simulated at the cell level with the Monte Carlo code SERPENT-1.1.7 with two modern data libraries, ENDF-B/VII and JEFF-3.1.1. The BFS-2 critical zero-power facility at the Inst. of Physics and Power Engineering (IPPE) was designed for simulations of the core and shielding of sodium-cooled, fast reactors, for neutron data validation and comparison with experimental results. At the BFS-2 facility, the BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor, with hybrid MOX fuel and stainless steel reflectors. A UO{sub 2} blanket and a large non-homogeneous stainless-steel reflector surround the core. The lattice is hexagonal of pitch 5.1 cm and metallic dowels are used to keep in central position cylindrical rods made of different types of material (fissile, fertile, blanket, plenum, shielding and absorber). A typical subassembly is formed in piling up various pellets of about 1 cm in height and 4.6 cm in diameter, conferring large heterogeneity in the axial direction. The full-core model development was a complex task due to the large number of subassemblies and the axial subassembly heterogeneity. In ERANOS-2.2, it was necessary to homogenize axially per region the pellets used to form the subassembly. The self-shielded macroscopic cross-sections were calculated using the cell code ECCO in association with JEFF-3.1 and ENDF/B-VI.8 data libraries. The core calculations were performed with broad cross-sections data in 33 neutron energy groups with the solver AVNM in the diffusion approximation, mostly. In MCNPX-2.7.0, a step-by-step approach was used, starting with a model in which the fissile rods were simulated on a homogeneous level, to finally integrate the actual heterogeneous description of the subassemblies. The code-to-code cell analysis performed between ECCO, SERPENT and MCNPX with different modern nuclear data library revealed that the results for the infinite multiplication factor between Monte Carlo and deterministic analysis are in good agreement ({Delta}p < 100 pcm). The differences between the results were observed to be larger for the neutron data libraries, with reactivity differences up to 350 pcm. (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

Calculation of the Phenix end-of-life test in natural circulation with the CATHARE 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. One of the advantages pointed up for fast reactors cooled by heavy liquid metal is the possibility of decay heat removal based on natural convection. The promotion of this passive cooling mode in future safety demonstrations will involve the use of adapted and validated numerical codes. After the final shutdown of the Phenix sodium cooled fast reactor in 2009, a set of tests covering different areas was conducted for code validation, including a natural circulation test in the primary circuit. Experimental data were issued by CEA to organize a benchmark exercise in the frame of an IAEA Coordinated Research Project (CRP), with the objective to assess the system-codes capability in simulating the thermal-hydraulics behavior of sodium cooled fast reactors in such accidental conditions. IRSN participated to this benchmark with the CATHARE code. This code, co-developed by CEA, EDF, AREVA and IRSN and widely used for PWR safety studies, was recently extended for sodium applications. This paper presents the CATHARE modeling of the Phenix primary circuit and the results obtained. A relatively good agreement was found with available measurements considering the experimental uncertainties. This work stressed the local aspects of phenomena occurring during the natural convection establishment and the limits of a 0D/1D approach. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Experiment on large scale plume interaction with a stratified gas environment resembling the thermal activity of a autocatalytic recombiner

Computational Fluid Dynamics codes (CFD) are increasingly being used to simulate containment conditions after various transient accident scenarios. Consequently, the reliability of such codes must be tested against experimental data. Such validation experiments related to gas mixing and hydrogen transport within containment compartments addressing the effect of heat source are presented in this paper. The experiments were conducted in the large-scale thermal-hydraulics PANDA facility located at the Paul-Scherrer-Inst. (PSI) in Switzerland, in the frame of the OECD/SETH-2 project. A 10 kW electric heater simulating the thermal activity of the autocatalytic recombiner was activated at full power in a containment vessel at the top of which a thick helium layer is initially present. The hot plume interacts with the bottom of the helium layer which is slowly eroded until complete break up at 1350 s. After final erosion of the layer a strong temperature and concentration gradient is maintained in the vessel below the heater inlet as well as in the adjacent vessel below the interconnecting pipe. A detailed characterization of the operating heater suggests the presence of cold gas ingress at the outlet that affects the flow in the chimney. This can be of concern if present in a real PAR unit. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗