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Accepted Manuscripts

BASIC VIEW  |  EXPANDED VIEW
research-article  
Huseyin Atilla Ozgener and Ceyhun Yavuz
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4042021
An analytical solution of the two-group diffusion equations is derived for multiregion source-driven subcritical systems in spherical geometry. An analytical formulation for the calculation of the effective multiplication factor is also presented. Using typical two-group cross sections characterizing source, buffer and blanket regions, parameters such as neutron amplification, source efficiency and blanket fast flux peaking factor are calculated. The criticality solution is utilized to calculate the effective multiplication factor and the neutron source efficiency.The dependency of the calculated global parameters on design variables like the source, buffer, blanket thicknesses and subcriticality level is studied. Thin source regions result in very high neutron amplification, at the expense of high blanket fast flux peaking factors. If a buffer region is put between the source and the blanket regions, flux peaking could be reduced at the expense of reduced neutron amplification. If the subcriticality level can be reduced without jeopardizing safety, the neutron amplification increases and the fast flux peaking is reduced.
TOPICS: Diffusion (Physics), Neutrons, Polishing equipment, Cross section (Physics), Design, Geometry, Neutron sources, Safety
research-article  
Jaroslav Kotowski, Tom S Cernousek, Filip Jankovsk, Pavel Kus, Petr Pol vka, Martin Skala, Hana Kov rov and Milan Zuna
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4041790
Granite block, acquired from a quarry Pansk Dubenka located in the Czech Republic and used in presented experiments, is part of the same bedrock that can be potentially used for a deep geological repository. It is important to characterize advection in fractured rock to assess possible groundwater contamination. Newly used method - 3D scanning using Hexagon Romer Arm was implemented to characterize the morphology of an examined fractured block with a crevice. The scanning technology provides the possibility to digitalize the rock surface. The scanning can be also used to determine any changes in the rock surface. The block was instrumented by tubing and the crevice was sealed using a silicone. Flow paths were investigated by the comparison of fluid weights on the outlet on every output/site. The Hexagon Romer Arm is an ideal tool for the precise determination of a crevice's width in its full volume.
TOPICS: Flow (Dynamics), Fluids, Tubing, Contamination, Instrumentation, Rocks, Silicones, Groundwater
research-article  
Tetsuaki Takeda
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4041690
When a depressurization accident of a Very-High-Temperature Reactor (VHTR) occurs, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Therefore, in order to predict or analyze the air ingress phenomena during a depressurization accident, it is important to develop a method for the prevention of air ingress during an accident. In particular, it is also important to examine the influence of localized natural convection and molecular diffusion on the mixing process from a safety viewpoint. Experiments and numerical analysis using three-dimensional (3D) CFD code have been carried out to obtain the mixing process of two-component gases and the flow characteristics of localized natural convection. The numerical model consists of a storage tank and a reverse U-shaped vertical rectangular passage. One sidewall of the high-temperature side vertical passage is heated and the other sidewall is cooled. The low-temperature vertical passage is cooled by ambient air. The storage tank is filled with heavy gas and the reverse U-shaped vertical passage is filled with a light gas. The results obtained from the experiments were quantitatively simulated using 3D numerical analysis. The two component gases were mixed via molecular diffusion and natural convection. After some time elapsed, natural circulation occurred through the reverse U-shaped vertical passage. These flow characteristics are the same as those of phenomena generated in the passage between a permanent reflector and a pressure vessel wall of the VHTR.
TOPICS: Fluids, Gases, Numerical analysis, Very high temperature reactors, Natural convection, Accidents, Flow (Dynamics), Diffusion (Physics), Storage tanks, Graphite, High temperature, Reactor vessels, Computational fluid dynamics, Low temperature, Safety, Computer simulation, Pressure vessels
research-article  
Tomás Czakoj and Evzen Losa
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4041692
Three dimensional Monte Carlo code KENO-VI of SCALE-6.2.2 code system was applied for criticality calculation of LR-0 reactor core. A central module placed in the center of the core was filled by graphite, FLINA and FLIBE materials. The multiplication factor was obtained for all cases using both ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. Obtained results were compared with benchmark calculations in MCNP6 using ENDF/B-VII.0 library. The results of KENO-VI calculations are found to be in good agreement with results obtained by the MCNP6. The discrepancies are typically within tens of pcm excluding the case with FLINA filling. Sensitivities and uncertainties of the reference case with no filling were determined by a continuos-energy version of TSUNAMI sequence of SCALE-6.2.2. The obtained uncertainty in multiplication factor due to the uncertainties in nuclear data is about 650 pcm with ENDF/B-VII.1.
TOPICS: Uncertainty, Tsunamis, Graphite
research-article  
Anna Hojna, Fosca Di Gabriele, Michal Chocholousek, Zbynek Spirit and Lucia Rozumova
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4041564
The austenitic steel 15-15Ti is being considered as one of the candidate materials for internal structural components of future Heavy Liquid Metal nuclear systems. This work studies the steel compatibility with liquid PbBi. Constant Extension Rate Tensile (CERT) tests of tapered specimens were used to study sensitivity to Liquid Metal Embrittlement (LME) and crack initiation. The taper creates a variation of stress along the gauge length which allows the identification of the stress and strain for the crack appearance. Testing was performed in air and in PbBi with 10-6 - 10-12 wt. % oxygen content at 300°C. Post-test observation by SEM highlighted the crack morphology. An evaluation of the environmental effect on the crack initiation is presented.
TOPICS: Steel, Fracture (Materials), Liquid metals, Stress, Gages, Structural elements (Construction), Testing, Embrittlement, Oxygen
research-article  
Valentyn Tsisar, Carsten Schroer, Olaf Wedemeyer, Aleksandr Skrypnik and Jürgen Konys
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4041432
Corrosion behavior of 9 %Cr F/M P92, E911 and EUROFER steels was investigated in flowing (2 m/s) Pb-Bi with 1E-7 mass%O at 450 and 550 °C for up to 8766 and 2011 h, respectively. The steels show mixed corrosion modes simultaneously revealing protective scale formation, accelerated oxidation and solution-based attack. At 450 °C, the accelerated oxidation resulted in a metal recession averaging 6 µm (± 2 µm) after ~ 8766 h while local solution-based corrosion attack ranged from ~40 to 350 µm. At 550 °C, the accelerated oxidation resulted in a metal recession of about 10 µm (± 2 µm) after ~ 2011 h. Solution-based corrosion attack appears more regularly at 550 °C, with a maximum depth ranged from ~90 to 1000 µm. Incubation time for solution based attack at 450 °C is 500-2000 h and ? 300 h at 550 °C.
TOPICS: Corrosion, Martensitic steel, Oxygen, oxidation, Metals, Steel, Incrustations
research-article  
Christopher J. Hurt, James D. Freels, Prashant K. Jain and Guillermo Maldonado
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4041295
Safety analyses at the High Flux Isotope Reactor (HFIR) are required to qualify irradiation of production targets containing neptunium dioxide/aluminum cermet (NpO2/Al) pellets for the production of plutonium-238 (238Pu). High heat generation rates due to a fertile starting material (237Np), low melting temperatures, and previously unstudied material irradiation behavior (i.e. swelling/densification, fission gas release) require a sophisticated set of steady-state thermal simulations in order to ensure sufficient safety margins. Experience gained from previous models for preliminary target designs is incorporated into a more comprehensive production target model designed to qualify a target for 3 cycles of irradiation and illuminate potential in-reactor behavior of the target.
TOPICS: Safety, Thermomechanics, Irradiation (Radiation exposure), Simulation, Melting, Engineering simulation, Cycles, Steady state, Heat, Temperature, Nuclear fission, Isotopes, Aluminum, Cermets
research-article  
Marcin Kopec and Martina Malá
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4041277
The ultrasonic measurements have a long history of utilization in the industry, also in the nuclear field. As the ultrasonic transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur - the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard post-irradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (towards the probe) and perpendicular (sideways against the probe) direction.
TOPICS: Ultrasonic measurement, Nuclear fuels, Fuel rods, Probes, Ultrasonic transducers, Inspection, Radiation (Physics), Particulate matter, Fuels, Safety, Manufacturing, Irradiation (Radiation exposure), Coolants, Cladding systems (Building), Nuclear reactors
Technical Brief  
Martin Schulc, Michal Košt'ál, Evžen Novák, Jan Šimon and Nicola Burianová
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4039774
The presented work deals with 23Na(n,2n)22Na and 127I(n,2n)126I reactions in the 252Cf spontaneous fission neutron source. 252Cf neutron source with approximate emission of 6E8 n/s was employed for the irradiation of sodium iodide. The spectrum averaged cross sections were then inferred from experimentally determined reaction rates and compared with calculations in MCNP6 using various nuclear data libraries. The experimental reaction rates were derived from the Net Peak Areas measured using the high purity germanium spectroscopy. The measured spectrum averaged cross section for the 23Na(n,2n)22Na reaction in the 252Cf spectrum was determined as equal to 8.98 ± 0.32 µb. The resulting spectrum averaged cross section for the 127I(n,2n)126I reaction in the 252Cf spectrum was derived as 2.044 ± 0.0072 mb. These experimental data can be used for nuclear data libraries validation and to specify high energy tail of the 252Cf neutron spectrum.
TOPICS: Neutron sources, Chemical kinetics, Spectroscopy, Irradiation (Radiation exposure), Cross section (Physics), Germanium, Sodium, Emissions, Nuclear fission, Neutrons
research-article  
Alex Matev
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036986
Many researchers are investigating the potential of lead-bismuth cooled fast reactors for producing electricity, as well as for the safe transmutation of minor actinides and the nuclear incineration of long-lived fission products. The paper presents the results from simulating with the RELAP5-3D code of natural circulation in a generic design of a pool-type nuclear reactor with lead-bismuth eutectic alloy (LBEA) as a primary, and water/steam as a secondary coolant. The simulation results provide valuable insights in the evolution of key reactor safety-relevant phenomena and support also the qualified use of system analysis codes as RELAP5-3D for the simulation of transients in pool-type reactor systems.
TOPICS: Simulation, Coolants, Eutectic alloys, Fast neutron reactors, Nuclear reactors, Simulation results, Steam, Water, Transients (Dynamics), Design, Nuclear fission, Systems analysis, Safety

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