| 1. |
- Bechta, Sevostian, et al.
(författare)
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INTERACTION BETWEEN MOLTEN CORIUM UO2+x-ZrO2-FeOy AND VVER VESSEL STEEL
- 2010
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Ingår i: Nuclear Technology. - : American Nuclear Society. - 0029-5450 .- 1943-7471. ; 170:1, s. 210-218
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Tidskriftsartikel (refereegranskat)abstract
- In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO2+x-ZrO2-FeOy melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction zone. The available experimental data have been used to develop a correlation for the corrosion rate as afunction of temperature and heat flux.
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| 2. |
- Dessirier, Benoît, et al.
(författare)
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Modeling Two-Phase-Flow Interactions across a Bentonite Clay and Fractured Rock Interface
- 2014
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 187:2, s. 147-157
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Tidskriftsartikel (refereegranskat)abstract
- Deep geological repositories are generally considered as suitable environments for final disposal of spent nuclear fuel. In the Swedish and Finnish repository design concept, canisters are to be placed in deep underground tunnels in sparsely fractured crystalline bedrock, in deposition holes in which each canister is embedded with an expansive bentonite-clay-mixture buffer. A set of semigeneric two-dimensional radially symmetric TOUGH2 simulations are conducted to investigate the multiphase dynamics and interactions between water and air in a bentonite-rock environment. The main objective is to identify how sensitive saturation times of bentonite are to the geometry of the rock fractures and to commonly adopted simplifications in the unsaturated flow description such as Richards assumptions. Results show that the location of the intersection between the fracture system and the deposition hole is a key factor affecting saturation times. A potential long-lasting desaturation of the rock matrix close to the bentonite-rock interface is also identified extending up to 10 cm inside the rock. Two-phase-flow models predict systematically longer saturation times compared to a simplified Richards approximation, which is frequently used to represent unsaturated flows. The discrepancy diverges considerably as full saturation is approached.
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| 3. |
- Dykin, Victor, 1985, et al.
(författare)
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Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors
- 2012
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 183:3, s. 354-366
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Konferensbidrag (refereegranskat)abstract
- This paper reports on the development and application of a method of emulating bubbly flow by generating bubbles with random sampling methods. The purpose of the modeling is that by using the simulated random two phase flow as input, one can generate "synthetic" neutron noise signals by convoluting the input with a simplified neuronic transfer function, on which the possibility of reconstructing the axial void profile from in-core neutron noise measurements can be studied by standard spectral noise analysis methods. The long term goal of this work is to elaborate methods of neutron noise analysis, by which the local void fraction in a boiling water reactor can be determined by measurements. In this preliminary stage, two methods for the reconstruction of the axial void and the velocity profiles are discussed. The first method is based on the break frequency of the neutron auto-power spectrum, whereas the second method only utilizes the information in the transit time of the void fluctuations between axial pairs of neutron detectors. A clear and monotonic relationship between the chosen observables and the two-phase flow properties was found, but an accurate determination of the void fraction requires further development and testing of the various unfolding alternatives.
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| 4. |
- Gajev, Ivan, et al.
(författare)
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Sensitivity and Uncertainty of OECD Benchmark Ringhals-1TRACE/PARCS Stability Prediction
- 2012
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 180:3, s. 383-398
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Tidskriftsartikel (refereegranskat)abstract
- Unstable behavior of boiling water reactors (BWRs) is known to occur during operation at certain power and flow conditions. This paper reports on an uncertainty study of the impact of various parameters on the prediction of the stability of the BWR within the framework of the Organisation for Economic Co-operation and Development Ringhals Unit 1 (Ringhals-1) Stability Benchmark. The time domain code TRACE/PARCS was used in the analysis. The paper is divided into two parts: a sensitivity study on numerical parameters (nodalization, time step, etc.) and an uncertainty analysis of the stability event. The sensitivity study was based on a space-time converged solution, and the most important neutronic and thermal-hydraulic parameters were identified for parameterization. The uncertainty calculation was then performed using the well-established propagation of input errors methodology. Finally, the Spearman Rank method was used to identify the most influential parameters affecting the stability of Ringhals-1.
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| 5. |
- Grape, Sophie, et al.
(författare)
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Partial Defect Evaluation Methodology for Nuclear Safeguards Inspections of Used Nuclear Fuel Using the Digital Cherenkov Viewing Device
- 2014
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 186:1, s. 90-98
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Tidskriftsartikel (refereegranskat)abstract
- This paper describes possible ways of analyzing and interpreting data obtained using the digital Cherenkov viewing device on spent nuclear fuel assemblies for the identification of partial defects in the fuel. According to the terminology of the International Atomic Energy Agency, partial defects refer to items, for instance, fuel assemblies, that are manipulated to the extent that a fraction of the fuel material is diverted or substituted. Analysis can be performed either by using a measure of the total light intensity or by identifying the light distribution pattern emanating from the spent nuclear fuel, the goal of either type of analysis being a quantitative measure that can be used in the data interpretation step. Two possible data interpretation alternatives are presented here: the threshold method and the hypothesis testing method. This paper summarizes some of the simulation studies and results that have been obtained, related to the two analysis and data interpretation methodologies.
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| 6. |
- Holcombe, Scott, et al.
(författare)
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Method For Analyzing Fission Gas Release In Fuel Rods Based On Gamma-Ray Measurements Of Short-Lived Fission Products
- 2013
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 184:1, s. 96-106
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Tidskriftsartikel (refereegranskat)abstract
- Fission gases are produced as a result of fission reactions in nuclear fuel. Most of these gases remain trapped within the fuel pellets, but some may be released to the fuel rod internal gas volume under certain conditions. This phenomenon of fission gas release is important for fuel performance since the released gases can degrade the thennal properties of the fuel rod. fill gas and contribute to increasing fuel rod internal pressure. Various destructive and nondestructive methods are available for determining the amount of fission gas release; however, the current methods are primarily useful for determining the integrated fission gas release fraction, i.e., the amount of fission gas produced in the fuel that has been released to the free rod volume over the entire lifetime of a nuclear fuel rod. In this work, a method is proposed for determining the fission gas release that occurs during short irradia-tion sequences. The proposed method is based on spectroscopic measurements of gamma rays emitted in the decay of short-lived fission gas isotopes. Determining such sequence-specific fission gas release can be of interest when evaluating the fuel behavior for selected times during irradiation, such as during power ramps. The data obtained in this type of measurement may also be useful for investigating the mechanisms behind fission gas release for fuel at high burnup. The method is demonstrated based on the analysis of experimental gamma-ray spectra previously collected using equipment not dedicated for this purpose; however, the analysis indicates the feasibility of the method. Further evaluation of the method is planned, using dedicated equipment at the Halden Boiling Water Reactor.
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| 7. |
- Kozlowski, Tomasz, et al.
(författare)
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QUALIFICATION OF THE RELAP5/PARCS CODE FOR BWR STABILITY EVENTS PREDICTION
- 2011
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 174:1, s. 51-63
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Tidskriftsartikel (refereegranskat)abstract
- The present study is concerned with the capability of a coupled neutron-kinetic/thermal-hydraulic code system RELAP5/PARCS for the numerical prediction of global core stability condition and instability transients. The work is motivated by the need to assess the safety significance of a number of stability transients that trigger core instability and challenge reactor protection systems. The technical approach adopted is done both to learn from real stability events and to perform analysis of idealized well-defined transients in a real plant and core configuration. In this paper, we show that the code system can serve as a unique and powerful tool to provide a consistent and reasonably reliable prediction of stability boundary even in complex plant transients. However, the prediction quality of the instability transients, i.e., core behavior without scram namely, parameters of the limit cycle remains questionable. We identify, two main factors for future studies (two-phase flow regimes in oscillatory flow and algorithm for effective grouping of thermal-hydraulic channels) as key to enhancing the predictive capability of the existing coupled code system for boiling water reactor stability.
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| 8. |
- Kudinov, Pavel, et al.
(författare)
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THE DEFOR-S EXPERIMENTAL STUDY OF DEBRIS FORMATION WITH CORIUM SIMULANT MATERIALS
- 2010
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 170:1, s. 219-230
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Tidskriftsartikel (refereegranskat)abstract
- Characteristics of corium debris beds formed in a severe core melt accident are studied in the Debris Bed Formation-Snapshot (DEFOR-S) test campaign, in which superheated binary-oxidic melts (both eutectic and non-eutectic compositions) as the corium simulants are discharged into a water pool. Water subcooling and pool depth are found to significantly influence the debris fragments' morphology and agglomeration. When particle agglomeration is absent, the tests produced debris beds with porosity of similar to 60 to 70%. This porosity is significantly higher than the similar to 40% porosity broadly used in contemporary analysis of corium debris coolability in light water reactor severe accidents. The impact of debris formation on corium coolability is further complicated by debris fragments' sharp edges, roughened surfaces, and cavities that are partially or fully encapsulated within the debris fragments. These observations are made consistently in both the DEFOR-S experiments and other tests with prototypic and simulant corium melts. Synthesis of the debris fragments from the DEFOR-S tests conducted under different melt and coolant conditions reveal trends in particle size, particle sphericity, surface roughness, sharp edges, and internal porosity as functions of water subcooling and melt composition. Qualitative analysis and discussion reaffirm the complex interplay between contributing processes (droplet interfacial instability and breakup, droplet cooling and solidification, cavity formation and solid fracture) on particle morphology and, consequently, on the characteristics of the debris beds.
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| 9. |
- Li, Liangxing, et al.
(författare)
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Experimental Study of Two-Phase Flow Regime and Pressure Drop in a Particulate Bed Packed with Multidiameter Particles
- 2012
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Ingår i: Nuclear Technology. - : American Nuclear Society. - 0029-5450 .- 1943-7471. ; 177:1, s. 107-118
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Tidskriftsartikel (refereegranskat)abstract
- This paper documents an experimental study on two-phase flow regimes and frictional pressure drop characteristics in a particulate (porous) bed packed with multidiameter (1.5-, 3-, and 6-mm) glass spheres. The experimental results provide new data to validate/develop hydrodynamic models for coolability analysis of debris beds formed in fuel-coolant interactions during a postulated severe accident. The POMECO-FL test facility is employed to perform the experiment, with the spheres packed in a test section of 90 mm diameter and 635 mm height. The pressure drops are measured for air/water two-phase flow through the packed bed, and flow patterns are obtained by means of visual observations. Meanwhile, local void fraction in the center of the bed is measured by a microconductive probe.The experimental results show that the frictional pressure drop of single-phase flow through the bed can be predicted by the Ergun equation, if the area mean diameter of the particles is chosen in the calculation. Given the so-determined effective particle diameter, the estimation of the Reed model for two-phase flow pressure gradient in the bed has a good agreement with the experimental data. The characteristics of the local void fraction can be used to predict flow pattern and mean void fraction. It is observed that slug flow prevails when the mean void fraction is <0.5, whereas annular flow dominates after the mean void fraction is >0.7. If the effective particle diameter is further used as an influential parameter in flow pattern identification, the observed flow regimes of two-phase flow in porous media are well predicted by the existing flow pattern map.
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| 10. |
- Loberg, John, 1980-, et al.
(författare)
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Homogenization of Cross Sections and Computation of Discontinuity Factors for a Real 3D BWR Bottom Reflector for Comparison with Lattice and Nodal Codes
- 2012
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 177:1, s. 1-7
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Tidskriftsartikel (refereegranskat)abstract
- Boiling water reactor (BWR) bottom reflector calculations in lattice codes such as CASMO are presently used only to produce accurate boundary conditions for core interfaces in nodal diffusion codes. Homogenized cross-section constants and discontinuity factors are calculated in one dimension (1-D) without the explicit presence of the control rod absorber. If the spatial flux in a BWR bottom reflector is required, for example, for depletion calculations of withdrawn control rods, the homogenization of the reflector must be based on a representation of the three-dimensional (3-D) geometry and material composition that is as true as possible. This paper investigates differences in cross-section and discontinuity factors from 1-D calculations in CASMO with 3-D Monte Carlo calculations of a realistic bottom reflector model in MCNP5. The cross-section and discontinuity factors from CASMO and MCNP5 are furthermore implemented in the nodal diffusion code SIMULATES to investigate the effect on the neutron fluxes in the bottom reflector. The results show that for the case investigated, the 1-D homogenization in CASMO5 produces a 26% overestimation of the homogenized thermal absorption cross section in the reflector and a 62% underestimation of the homogenized fast absorption cross section. These cross-section differences have essentially no impact on the neutron flux in the core but cause a 4.5% and 12.3% underestimation of the thermal and fast neutron flux, respectively, in the reflector region.
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