| 1. |
- Arzhanov, Vasily, et al.
(författare)
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Localization of a vibrating control rod pin in pressurized water reactors using the neutron flux and current noise
- 2000
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 131:2, s. 239-251
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Tidskriftsartikel (refereegranskat)abstract
- It has been proposed that the fluctuations of the neutron current called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The possibility of the localization of a vibrating control rod pin in a pressurized water reactor control assembly is investigated by using the scalar neutron noise and the two-dimensional radial current noise as measured at one central point in the assembly. Art explicit localization technique is elaborated in which the searched position is determined as the absolute minimum of a minimization function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method.
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| 2. |
- Bechta, Sevostian, et al.
(författare)
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INTERACTION BETWEEN MOLTEN CORIUM UO2+x-ZrO2-FeOy AND VVER VESSEL STEEL
- 2010
-
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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| 3. |
- Bengtsson, Martin, et al.
(författare)
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Experimental method for verification of calculated Cs-137 content in nuclear fuel assemblies
- 2022
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Ingår i: Nuclear Technology. - : Informa UK Limited. - 0029-5450 .- 1943-7471. ; 208:2, s. 295-302
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Tidskriftsartikel (refereegranskat)abstract
- A method to determine the absolute activity of 137Cs in irradiated nuclear fuel is presented. Using a well-known point-like calibration source in combination with measurements of the gamma-ray intensity from the nuclear fuel and Monte Carlo calculations based on the nominal measurement geometry, the activity content can be determined without prior knowledge of the intrinsic detection efficiency of the gamma-ray detector. The presented method is tested using measurements of the 137Cs intensity from spent nuclear fuel of the pressurized water type at the central interim storage in Sweden. Using an assumption of homogeneous distribution of 137Cs throughout the fuel, we demonstrate a linear relationship between measured activity and the activity calculated by a state-of-the-art simulation code. For future studies, we suggest some factors that potentially can decrease the uncertainty in the correlation between measured and calculated activity.
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| 4. |
- Cholewa, W, et al.
(författare)
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Identification of loss-of-coolant accidents in LWRs by inverse models
- 2004
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 147:2, s. 216-226
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Tidskriftsartikel (refereegranskat)abstract
- This paper describes a novel diagnostic method based on inverse models that could be applied to identification of transients and accidents in nuclear power plants. In particular, it is shown that such models could be successfully applied to identification of loss-of-coolant accidents (LOCAs). This is demonstrated for LOCA scenarios for a boiling water reactor. Two classes of inverse models are discussed: local models valid only in a selected neighborhood of an unknown element in the data set, representing a state of a considered object, and global models, in the form of partially unilateral models, valid over the whole learning data set. An interesting and useful property of local inverse models is that they can be considered as example based models, i.e., models that are spanned on particular sets of pattern data. It is concluded that the optimal diagnostic method should combine the advantages of both models, i.e., the high quality of results obtained from a local inverse model and the information about the confidence interval for the expected output provided by a partially unilateral model.
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| 5. |
- De Geer, Lars-Erik, et al.
(författare)
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A Nuclear Jet at Chernobyl Around 21:23:45 UTC on April 25, 1986
- 2018
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Ingår i: Nuclear Technology. - : Informa UK Limited. - 0029-5450 .- 1943-7471. ; 201:1, s. 11-22
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Tidskriftsartikel (refereegranskat)abstract
- The nature of two explosions that were witnessed within 3 s at the Chernobyl-4 reactor less than a minute after 21:23:00 UTC on April 25, 1986, have since then been the subject of sprawling interpretations. This paper renders the following hypothesis. The first explosion consisted of thermal neutron mediated nuclear explosions in one or rather a few fuel channels, which caused a jet of debris that reached an altitude of some 2500 to 3000 m. The second explosion would then have been the steam explosion most experts believe was the first one. The solid support for this new scenario rests on two pillars and three pieces of corroborating evidence. The first pillar is that a group at the V. G. Khlopin Radium Institute in then Leningrad on April 29, 1986, detected newly produced, or fresh, xenon fission products at Cherepovets, 370 km north of Moscow and far away from the major track of Chernobyl debris ejected by the steam explosion and subsequent fires. The second pillar is built on state-of-the-art meteorological dispersion calculations, which show that the fresh xenon signature observed at Cherepovets was only possible if the injection altitude of the fresh debris was considerably higher than that of the bulk reactor core releases that turned toward Scandinavia and central Europe. These two strong pieces of evidence are corroborated by what were manifest physical effects of a downward jet in the southeastern part of the reactor, by seismic measurements some 100 km west of the reactor, and by observations of a blue flash above the reactor a few seconds after the first explosion.
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| 6. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Evaluation of the boron dilution method for Moderator Temperature Coefficient measurements
- 2002
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 140:1, s. 147-163
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Tidskriftsartikel (refereegranskat)abstract
- A measurement of the at-power moderator temperature coefficient (MTC) at the pressurized water reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed. The measurement was performed when the boron concentration decreased under 300 ppm in the reactor coolant system, by using the boron dilution method. Detailed calculations were made to estimate all reactivity effects taking place during such a measurement. These effects can only be accounted for through static core calculations that allow calculating contributions to the reactivity change induced by the moderator temperature change. All the calculations were performed with the Studsvik Scandpower SIMULATE-3 code. Analysis of the measurement showed that the contribution of the Doppler effect (in the fuel) was almost negligible, whereas the reactivity effects due to other than the Doppler fuel coefficient and the boron change were surprisingly significant. It was concluded that due to the experimental inaccuracies, the uncertainty associated with the boron dilution method could be much larger than previously expected. The MTC might then be close to -72 pcm/oC, whereas the main goal of the measurement is to verify that the MTC is larger (less negative) than this threshold. The usefulness of the boron dilution method for MTC measurements can therefore be questioned.
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| 7. |
- 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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| 8. |
- 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
-
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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| 9. |
- Dykin, Victor, 1985, et al.
(författare)
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The Molten Salt Reactor Point-Kinetic Component of Neutron Noise in Two-Group Diffusion Theory
- 2016
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 193:3, s. 404-415
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Tidskriftsartikel (refereegranskat)abstract
- The derivation of the point-kinetic component of the neutron noise in two-group diffusion theory in molten salt reactors (MSRs), based on different techniques, is discussed. First, the point-kinetic component is calculated by projecting the corresponding full space-frequency-dependent solution onto the static adjoint. Then, following the standard procedure in reactor physics, the point-kinetic solution is determined by solving the linearized point-kinetic equations. Both results are thereafter analyzed and compared quantitatively. Such a comparison clearly indicates that the solution obtained by the conventional derivation, i.e., from the point-kinetic equations, significantly differs from the exact one and is not able to reproduce certain features of the latter. Similar discrepancies between the two methods were also pointed out and confirmed earlier in one-group MSR calculations.
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| 10. |
- Eriksson, Marcus, et al.
(författare)
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Inherent Safety of Fuels for Accelerator-driven Systems
- 2005
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 151:3, s. 314-333
-
Tidskriftsartikel (refereegranskat)abstract
- Transient safety characteristics of accelerator-driven systems using advanced minor actinide fuels have been investigated. Results for a molybdenum-based Ceramic-Metal (CerMet) fuel, a magnesia-based Ceramic-Ceramic fuel, and a zirconium-nitride-based fuel are reported. The focus is on the inherent safety aspects of core design. Accident analyses are carried out for the response to unprotected loss-of-flow and accelerator beam-overpower transients and coolant voiding scenarios. An attempt is made to establish basic design limits for the fuel and cladding. Maximum temperatures during transients are determined and compared with design limits. Reactivity effects associated with coolant void, fuel and structural expansion, and cladding relocation are investigated. Design studies encompass variations in lattice pitch and pin diameter. Critical mass studies are performed. The studies indicate favorable inherent safety features of the CerMet fuel. Major consideration is given to the potential threat of coolant voiding in accelerator-driven design proposals. Results for a transient test case study of a postulated steam generator tube rupture event leading to extensive coolant voiding are presented. The study underlines the importance of having a low coolant void reactivity value in a lead-bismuth system despite the high boiling temperature of the coolant. It was found that the power rise following a voiding transient increases dramatically near the critical state. The studies suggest that a reactivity margin of a few dollars in the voided state is sufficient to permit significant reactivity insertions.
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| 11. |
- 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
-
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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| 12. |
- Gottlieb, C., et al.
(författare)
-
Feasibility study on transient identification in nuclear power plants using support vector machines
- 2006
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 155:1, s. 67-77
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Tidskriftsartikel (refereegranskat)abstract
- Support vector machines (SVMs), a relatively new paradigm in statistical learning theory, are studied for their potential to recognize transient behavior of detector signals corresponding to various accident events at nuclear power plants (NPPs). Transient classification is a major task for any computer-aided system for recognition of various malfunctions. The ability to identify the state of operation or events occurring at an NPP is crucial so that personnel can select adequate response actions. The Modular Accident Analysis Program, version 4 (MAAP4) is a program that can be used to model various normal and abnormal events in an NPP. This study uses MAAP signals describing various loss-of-coolant accidents in boiling water reactors. The simulated sensor readings corresponding to these events have been used to train and test SVM classifiers. SVM calculations have demonstrated that they can produce classifiers with good generalization ability for our data. This in, turn indicates that SVMs show promise as classifiers for the learning problem of identifying transients.
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| 13. |
- 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
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 186:1, s. 90-98
-
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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| 14. |
- Hansson, Roberta Concilio, et al.
(författare)
-
Dynamics and preconditioning in a single-droplet vapor explosion
- 2009
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 167:1, s. 223-234
-
Tidskriftsartikel (refereegranskat)abstract
- The present study aims to develop a mechanistic understanding of the thermal-hydraulic processes in a vapor explosion, which may occur in nuclear power plants during a hypothetical severe accident, involving interactions of high-temperature corium melt and volatile coolant. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) were investigated in the Micro-Interactions in Steam Explosion Experiments (MISTEE) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematography and continuous X-ray radiography, called Simultaneous High-speed Acquisition of X-ray Radiography and Photography (SHARP). After an elaborate image processing, the SHARP images depict the evolution of both melt material (dispersal) and coolant (bubble dynamics) and their microscale interactions. The analysis of the data shows a deficiency in using the bubble dynamics alone to provide a consistent explanation of the energetic behavior. In contrast, the SHARP data reveal a correlation between the droplet's dynamics in the bubble's first cycle and the energetics of the subsequent explosive evaporation in the bubble's second cycle. The finding provides a basis to suggest that a so-called melt-droplet preconditioning, i.e., deformation/prefragmentation of a hot melt droplet immediately following the pressure trigger, is instrumental to the subsequent coolant entrainment, evaporation, and energetics of the resulting vapor explosion.
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| 15. |
- 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
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 184:1, s. 96-106
-
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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| 16. |
- Hultqvist, Martha, 1983-, et al.
(författare)
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Secondary doses in anthropomorphic phantoms irradiated with light ion beams
- 2009
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 168:1, s. 123-127
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Tidskriftsartikel (refereegranskat)abstract
- The mathematical anthropomorphic phantoms EVA-HIT and ADAM-HIT have been used in the Monte Carlo code SHIELD-HIT07 for simulations of lung tumor and prostate irradiation with light ions. Calculations were performed for 1H, 7Li, and 12C beams of energies in the range of 80 to 330 MeV/u. The secondary doses to organs, due to scattered primary ions and secondary particles produced in the phantoms, were studied taking into account the contribution from secondary neutrons, secondary protons, pions, and heavier fragments from helium to calcium. The doses to organs per dose to target (tumor) are of the order of 10-6 to 10-1 mGy Gy-1 and decrease with increasing distance from the target. In general the organ dose per target dose increases with increasing Z of the primary particle; however, for lighter primary ions (Z 3) and for organs close to the target, scattered primary particles show a nonnegligible dose contribution.
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| 17. |
- Hursin, Mathieu, et al.
(författare)
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Measurement of the Gas Velocity in a Water-Air Mixture in CROCUS Using Neutron Noise Techniques
- 2020
-
Ingår i: Nuclear Technology. - : Informa UK Limited. - 0029-5450 .- 1943-7471. ; 206:10, s. 1566-1583
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Tidskriftsartikel (refereegranskat)abstract
- The possibility of measuring the gas-phase velocity in a two-phase mixture through the use of neutron noise techniques is demonstrated in the zero-power reactor CROCUS of the Ecole Polytechnique Federale de Lausanne. It is the first step toward the experimental validation of an existing theoretical model whose objective is the reconstruction of the void profile in a channel. The use of zero-power research reactors is advantageous due to their clean environment in terms of signal fluctuations. To this end, a channel was installed in the reflector of CROCUS. A two-component mixture is generated inside the channel through the injection of compressed air. The signal fluctuations of neutron detectors located at various axial locations next to the channel are processed to determine the transit time of the gas phase between detectors. Four methods are presented based on the detector signal time series either in the time domain (time correlations between signals) or in the frequency domain (phase of the cross-power spectral density. All four methods returned consistent transit times and similar experimental uncertainty. The largest possible gas injection rates as well as the highest possible neutron flux level improve the visibility of the traveling perturbation and reduce the experimental uncertainty on the transit time for a given acquisition time. © 2020, © 2020 American Nuclear Society.
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| 18. |
- Jacobsson, Staffan, et al.
(författare)
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A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel Assemblies - II: Experimental Investigation
- 2001
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 135:2, s. 146-153
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Tidskriftsartikel (refereegranskat)abstract
- A tomographic method for verification of the integrity of used light water reactor fuel has been experimentally investigated. The method utilizes emitted gamma rays from fission products in the fuel rods. The radiation field is recorded in a large number of positions relative to the assembly, whereby the source distribution is reconstructed using a special-purpose reconstruction code.An 8 × 8 boiling water reactor fuel assembly has been measured at the Swedish interim storage (CLAB), using installed gamma-scanning equipment modified for the purpose of tomography. The equipment allows the mapping of the radiation field around a fuel assembly with the aid of a germanium detector fitted with a collimator with a vertical slit. Two gamma-ray energies were recorded: 662 keV (137Cs) and 1274 keV (154Eu). The intensities measured in 2520 detector positions were used as input for the tomographic reconstruction code. The results agreed very well with simulations and significantly revealed a position containing a water channel in the central part of the assembly.
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| 19. |
- Jansson, Peter, 1971-, et al.
(författare)
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Calculations of the Neutron Flux Outside BWR 8×8 Spent-Fuel Assemblies and the Sensitivity to Fuel Pin Diversion
- 2004
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 146:1, s. 58-64
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Tidskriftsartikel (refereegranskat)abstract
- The possibility of detecting replaced fuel rods in a spent-fuel assembly by means of measurement of the emitted neutron- and gamma-ray radiation has been investigated by computer simulations. The radiation field outside a boiling water reactor 8 × 8 fuel assembly with varying patterns of fuel rods replaced with lead dummies was calculated using a simple model for the source distribution and the Monte Carlo code MCNP-4C for the radiation field. In particular, the sensitivity of the thermal neutron field as measured in a Fork detector to various replacement patterns was investigated. The results suggest a detection limit of 5% of the fuel mass replaced, i.e., 3 out of 63 rods, independently of the pattern of the replaced rods.
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| 20. |
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| 21. |
- Jeltsov, Marti, 1985-, et al.
(författare)
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Parametric study of sloshing effects in the primary system of an isolated lead-cooled fast reactor
- 2015
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 190:1, s. 1-10
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Tidskriftsartikel (refereegranskat)abstract
- Risks related to sloshing of liquid metal coolant due to seismic excitation need to be investigated. Sloshing effects on reactor performance include first, fluid-structure interaction and second, gas entrapment in the coolant with subsequent transport of void to the core region. While the first can hypothetically lead to structural damage or coolant spill, the second increases the risk of a reactivity insertion accident and/or local dryout of the fuel. A two-dimensional computational fluid dynamics study is carried out in order to obtain insights into the modes of sloshing depending on the parameters of seismic excitation. The applicability and performance of the numerical mesh and the Eulerian volume of fluid method used to track the free surface are evaluated by modeling a simple dam break experiment. Sloshing in the cold plenum free surface region of the European Lead-cooled SYstem (ELSY) conceptual pool-type lead-cooled fast reactor (LFR) is studied. Various sinusoidal excitations are used to imitate the seismic response at the reactor level. The goal is to identify the domain of frequencies and magnitudes of the seismic response that can lead to loads threatening the structural integrity and possible core voiding due to sloshing. A map of sloshing modes has been developed to characterize the sloshing response as a function of excitation parameters. Pressure forces on vertical walls and the lid have been calculated. Finally, insight into coolant voiding has been provided.
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| 22. |
- Kaiserfeld, Thomas, et al.
(författare)
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Changing the System Culture : Mobilizing the Social Sciences in the Swedish Nuclear Waste System
- 2021
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Ingår i: Nuclear Technology. - : Informa UK Limited. - 0029-5450 .- 1943-7471. ; 207:9, s. 1456-1468
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Tidskriftsartikel (refereegranskat)abstract
- The purpose of this paper is to analyze how competence in the humanities and social sciences has been introduced into the system culture of the Swedish nuclear waste system (SNWS) traditionally dominated by scientists and engineers. In the spring of 1980, fierce local protests were directed against drilling teams sent out to investigate the geology of potential locations for a repository of spent nuclear fuel. This demonstrated the political and ethical dimensions of the waste issue and the limitations of the technocratic approach that had hitherto dominated the system culture of the SNWS. In order to counter this tendency, the government established an advisory board, Samrådsnämnden för kärnavfall (abbreviated KASAM), in 1985 with the task to widen the perspectives on the nuclear waste issue. KASAM engaged social scientists and humanists and started organizing annual workshops inviting engineers and scientists working with the waste issue to discuss its ethical and political dimensions. In the early 1990s, SKB, the Swedish implementer organization responsible for the management of nuclear waste, changed its strategy for finding suitable locations for a repository of spent nuclear fuel. Approval from the local population became a key condition. In the early 2000s, only two municipalities remained, both of them already housing nuclear power plants. After careful investigations and many deliberations, one of them was eventually chosen. The combination of KASAM’s activities to broaden the discussion and the local protests in many communities initiated a gradual change of the system culture within the SNWS. The initial technocratic approach was broadened to encompass ethical, social, and political aspects, and the main organizations now acknowledge that not only technical and scientific skills but also competence from social science and the humanities were of essence.
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| 23. |
- Kienzler, Bernhard, et al.
(författare)
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ACTINIDE MIGRATION IN FRACTURES OF GRANITE HOST ROCK : LABORATORY AND IN SITU INVESTIGATIONS
- 2009
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 165:2, s. 223-240
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Tidskriftsartikel (refereegranskat)abstract
- Within the scope of a cooperation between Svensk Karnbranslehantering AB and Forschungszentrum Karlsruhe, Institut fur Nukleare Entsorgung, a series of actinide migration experiments were performed both in the laboratory and at the Aspo Hard Rock Laboratory in Sweden. The objectives of these experiments were to quantify, the sorption of different actinide elements in single fractures of a granite host rock and to investigate the sorption mechanisms. To guarantee the most realistic conditions-as close to nature as possible-in situ experiments were performed in the Chemlab 2 borehole probe. These migration experiments were complemented by laboratory sorption and migration studies. The latter included batch experiments with flat chips of natural material extracted from fracture surfaces to identify the mineral phases relevant to radionuclide sorption by means of autoradiography. Scanning electron microscopy analyses provided information on the composition of sorption-relevant phases and X-ray photoelectron spectroscopy of Np, Tc, and Fe distribution revealed the redox states of these elements. Important mineral phases retaining all actinides and Tc were Fe-bearing phases. From the migration experiments, elution curves of the inert tracer (HTO), Np(V), U(VI), and to a small extent of Tc(VII) were obtained. Americium (III) and plutonium(IV) were not eluted. The mechanisms influencing the migration of the elements Np, U, and Tc depended on redox reactions. It was shown by various independent methods that Np(V) was reduced to the tetravalent state on the fracture surfaces, thus resulting in a pronounced dependence of the recovery on the residence time. Technetium was also retained in the tetravalent state. Elution of natural uranium from the granite drill cores was significant and is discussed in detail.
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| 24. |
- Kozlowski, Tomasz, et al.
(författare)
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Consistent comparison of the codes RELAP5/PARCS and TRAC-M/PARCS for the OECD MSLB coupled code benchmark
- 2004
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 146:1, s. 15-28
-
Tidskriftsartikel (refereegranskat)abstract
- A generalized interface module was developed for coupling any thermal-hydraulic code to any spatial kinetic code. In the design used here the thermal-hydraulic and spatial kinetic codes function as independent processes and communicate using the Parallel Virtual Machine software. This approach helps maximize flexibility while minimizing modifications to the respective codes. Using this interface, the U.S. Nuclear Regulatory Commission (NRC) three-dimensional neutron kinetic code, Purdue Advanced Reactor Core Simulator (PARCS), has been coupled to the NRC system analysis codes RELAP5 and Modernized Transient Reactor Analysis Code (TRAC-M). Consistent comparison of code results for the Organization for Economic Cooperation and Development/Nuclear Energy Agency main steam line break benchmark problem using RELAP5/PARCS and TRAC-M/PARCS was made to assess code performance.
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| 25. |
- Kozlowski, Tomasz, et al.
(författare)
-
QUALIFICATION OF THE RELAP5/PARCS CODE FOR BWR STABILITY EVENTS PREDICTION
- 2011
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 174:1, s. 51-63
-
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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| 26. |
- Kudinov, Pavel, et al.
(författare)
-
THE DEFOR-S EXPERIMENTAL STUDY OF DEBRIS FORMATION WITH CORIUM SIMULANT MATERIALS
- 2010
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 170:1, s. 219-230
-
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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| 27. |
- Li, Liangxing, et al.
(författare)
-
Experimental Study of Two-Phase Flow Regime and Pressure Drop in a Particulate Bed Packed with Multidiameter Particles
- 2012
-
Ingår i: Nuclear Technology. - : American Nuclear Society. - 0029-5450 .- 1943-7471. ; 177:1, s. 107-118
-
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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| 28. |
- Liu, J. S., et al.
(författare)
-
Effect of water radiolysis caused by dispersed radionuclides on oxidative dissolution of spent fuel in a final repository
- 2001
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 135:2, s. 154-161
-
Tidskriftsartikel (refereegranskat)abstract
- When released out of a canister, the radionuclides originally incorporated in the spent fuel can still deposit radiation energy (even more efficiently) into the pore water, cause water radiolysis, and produce oxidants in the buffering material. This phenomenon is termed secondary water radiolysis. The oxidants thus produced can possibly diffuse back to oxidize the spent fuel and to increase the oxidative dissolution rare of the fuel, The effect of the secondary water radiolysis has been identified and preliminarily addressed by a mass-balance model. To explore whether the effect is significant on spent-fuel dissolution, the upper-boundary limit of the effect has been set up by considering a scenario that is very unlikely to occur. Several extreme assumptions have been made: First, the canister fails completely 10(3) yr after deposition; second, the sl,ent fuel is oxidized instantaneously; and third, the radionuclides considered are those that dominantly contribute to radiolysis between 10(3) to 10(5) yr. With these assumptions, the spent-fuel dissolution rate can be increased dramatically if 10% or more of the oxidants produced by the secondary water radiolysis diffuse back to oxidize the spent fuel. It thus indicates that the effect of the secondary water radiolysis could be significant with some extreme assumptions. With more realistic assumptions, the effect could possibly become minimal. The subject is worth further investigation.
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| 29. |
- Liu, J. S., et al.
(författare)
-
Study of the consequences of secondary water radiolysis surrounding a defective canister
- 2003
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 142:3, s. 294-305
-
Tidskriftsartikel (refereegranskat)abstract
- In the concept of deep geological disposal of spent nuclear fuel, a chemically reducing environment in the near field of a repository is favorable for retaining the radionuclides in the fuel. Water radiolysis can possibly change a reducing environment in the near field to an oxidizing environment. In this paper, the consequences of secondary water radiolysis, caused by radionuclides released from the spent nuclear fuel and dispersed in the bentonite buffer surrounding a canister, have been studied. The canister is assumed to be initially defective with a hole of a few millimeters on its wall. The small hole will considerably restrict the transport of oxidants through the canister wall and the release of radionuclides to the outside of the canister. The spent fuel dissolution is assumed to be controlled by chemical kinetics at rates extrapolated from experimental studies. Two cases are considered. In the first case it is assumed that secondary phases of radionuclides [such as amorphous Pu(OH)(4) and AmOHCO3] do not precipitate inside the canister. The model results show that a relatively large domain of the near field can be oxidized by the oxidants of secondary radiolysis. In the second case it is assumed that secondary phases of radionuclides precipitate inside the canister, and the radionuclide concentration within the canister is controlled by its respective solubility limit. The amount of radionuclides released out of the canister will then be limited by the solubility of the secondary phases. The effect of the secondary radiolysis outside the canister on the rate of spent fuel oxidation inside a defective canister will be quite limited and can be neglected for any practical purposes in this case.
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| 30. |
- Liu, Longcheng, et al.
(författare)
-
A coupled model for oxidative dissolution of spent fuel and transport of radionuclides from an initially defective canister
- 2001
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 135, s. 273-285
-
Tidskriftsartikel (refereegranskat)abstract
- An earlier model for oxidative dissolution of spent fuel was developed by including the release behavior of actinides from the fuel surface and the barrier effect of Zircaloy claddings. The aim here is to explore the possibility and consequences of precipitation in the water film around the fuel pellets due to solubility, and transport limitations of nuclides. The model has been applied in the performance assessment of a damaged canister under natural repository conditions, by coupling to a redox-front-based model for transport of nuclides. The simulation results identify? that the time of penetration of the canister, the size of the damage, and the initial free volume of the fuel rods are important factors that dominate the dissolution behavior of the fuel matrix and thus the transport behavior of actinides in the nearfield of a repository.
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| 31. |
- Liu, Longcheng, et al.
(författare)
-
A reactive transport model for oxidative dissolution of spent fuel and release of nuclides within a defective canister
- 2002
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 137, s. 228-240
-
Tidskriftsartikel (refereegranskat)abstract
- In this study, we develop a mechanism-based model to take into account most of the important processes that may influence the dissolution behavior of spent fuel and subsequently the release behavior of nuclides within a defective canister in a final repository for high-level nuclear waste. The model is, in essence, a redox-controlled reactive transport model that provides a description of the mass transport of multiple species involved in both local equilibrium and kinetically controlled reactions in the system. The complexity of the kinetics of the various redox reactions involved and the requirement of the long-term prediction, however, make numerical implementation of the fully coupled model computationally inefficient. A series of scoping calculations was performed to highlight the local characteristics and behaviors of the system, and to provide a basis for refinement of the reactive transport model. The results indicate that the rapid buildup of hydrogen within the system is mainly attributed to corrosion of the cast-iron insert that primarily occurs under anaerobic conditions, rather than to radiolysis of water. The system that is rapidly in equilibrium with 50 bar hydrogen would then keep pH constant throughout the system. In addition, simulations suggest that reduction of dissolved hexavalent uranium by ferrous iron adsorbed onto the corrosion products and by dissolved H-2 are the most important mechanisms to retard the release of uranium out of the canister. More importantly, it is found that the pseudo stationary state approximation may well be applied to the system. This greatly simplifies the numerical implementation of the reactive transport model.
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| 32. |
- Liu, Longcheng, et al.
(författare)
-
Analysis of fluid flow and solute transport in a fracture intersecting a canister with variable aperture fractures and arbitrary intersection angles
- 2005
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 150:2, s. 132-144
-
Tidskriftsartikel (refereegranskat)abstract
- A multitude of simulations have been made for different types of rough-walled fractures, by using FEM-LAB((R)), to evaluate the mass transfer to and from water flowing through a fracture with spatially variable apertures and with an arbitrary angle of intersection to a canister that contains spent nuclear fuel. This paper presents and discusses only the results obtained for the Gaussian fractures. The simulations suggest that the intersection angle has only a minor influence on both the volumetric and the equivalent flow rates. The standard deviation of the distribution of the volumetric flow rates of the many realizations increases with increasing roughness and spatial correlation length of the aperture field, and so does that of the equivalent flow rates. The mean of the distribution of the volumetric flow rates is determined, however, solely by the hydraulic aperture, while that of the equivalent flow rates is determined by the mechanical aperture. Based upon the analytical solutions for the parallel plate model, it has been found that the distributions of both the volumetric and the equivalent flow rates are close to the Normal. Thus, two simple expressions can be devised to quantify the stochastic properties of fluid flow and solute transport through spatially variable fractures without making detailed calculations in every fracture intersecting a deposition hole or a tunnel.
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| 33. |
- Liu, Longcheng, et al.
(författare)
-
The effect of hydrogen on oxidative dissolution of spent fuel
- 2002
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 138, s. 69-78
-
Tidskriftsartikel (refereegranskat)abstract
- An earlier model for the oxidative dissolution of spent fuel is further developed by including the reductive effect of H-2, which is formed by both the radiolysis of ground-water and the anoxic corrosion of the cast iron insert of the canister. The kinetics of reduction of dissolved uranium species by dissolved hydrogen is derived from a series of previously published experimental studies. The simulation results suggest that the effect of autocatalytic reduction of hexavalent uranium by hydrogen may play an important role in controlling the dissolution of the fuel matrix within a canister. Further experimental studies are required to firmly verify these findings.
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| 34. |
- Loberg, John, 1980-, et al.
(författare)
-
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
-
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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| 35. |
- Marklund, Lars, et al.
(författare)
-
Impact of landscape topography and quaternary overburden on the performance of a geological repository of nuclear waste
- 2008
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 163, s. 165-179
-
Tidskriftsartikel (refereegranskat)abstract
- The topographical driving forces for groundwater on different spatial scales in several ways influence the performance of a repository for nuclear waste located at large depth in crystalline bedrock. We show that the relation between local topographical characteristics (topographical steepness and wavelengths) in the area of a repository (kilometer scale) and the large-scale (hundreds of kilometers) surroundings, together with repository depth, are the primary controls of residence time distributions and the discharge pattern of radionuclides released from an underground repository. In addition, the topography affects the groundwater flow at repository depth and, therefore, influences the long-time degradation of the repository. In the areas studied, all located in Sweden, the local topography mainly controls the groundwater flow down to a depth of ∼500 m, which is the suggested depth of the Swedish repository. The importance of the large-scale topography increases with depth but even at depth where local-scale topography is dominant, the continental-scale topography affects length and depth of flow paths as well as groundwater velocities. The impact of large-scale topography is particularly clear in areas where the steepness of local-scale landforms is relatively small. The study also shows that quaternary deposits (bedrock overburden) may have a significant impact on the overall residence times in the underground because of their hydraulic and sorption properties. This effect is fiirther enhanced by the fact that flow paths originating from repository depth generally emerge in topographical lows with relatively deep layers of quaternary deposits. The findings of this study underscore the need to consider multiscale topographical characteristics as well as bedrock overburden in assessments of radiological consequences of underground repositories.
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| 36. |
- Mickus, Ignas, et al.
(författare)
-
PERFORMANCE OF THE EXPLICIT EULER AND PREDICTOR-CORRECTOR-BASED COUPLING SCHEMES IN MONTE CARLO BURNUP CALCULATIONS OF FAST REACTORS
- 2015
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 191:2, s. 193-198
-
Tidskriftsartikel (refereegranskat)abstract
- We present a stability test of the explicit Euler and predictor-corrector based coupling schemes in Monte Carlo burnup calculations of the gas fast reactor fuel assembly. Previous studies have identified numerical instabilities of these coupling schemes in Monte Carlo burnup calculations of thermal-spectrum reactors due to spatial feedback induced neutron flux and nuclide density oscillations, where only sufficiently small time steps could guarantee acceptable precision. New results suggest that these instabilities are insignificant in fast-spectrum assembly burnup calculations, and the considered coupling schemes can therefore perform well in fast-spectrum reactor burnup calculations even with relatively large time steps. Note: Some figures in this technical note may be in color only in the electronic version.
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| 37. |
- Neretnieks, Ivars, et al.
(författare)
-
Density-Driven Mass Transfer in Repositories for Nuclear Waste
- 2019
-
Ingår i: Nuclear Technology. - : Taylor & Francis. - 0029-5450 .- 1943-7471. ; 205:6, s. 819-829
-
Tidskriftsartikel (refereegranskat)abstract
- In geologic repositories for nuclear waste located in crystalline rocks, the waste is surrounded by a bentonite buffer that in practice is not permeable to water flow. The nuclides must escape by molecular diffusion to enter the seeping water in the fractures of the rock. At high water-seepage rates, the nuclides can be carried away rapidly. The seepage rate of the water can be driven by the regional hydraulic gradient as well as by buoyancy-driven flow. The latter is induced by thermal circulation of the water by the heat produced by radionuclide decay. The circulation may also be induced by salt exchange between buffer and water in the fractures. The main aim of this paper is to explore how salt exchange between the backfill and mobile water in fractures, by buoyancy effects, can increase the escape rate of radionuclides from a repository. A simple analytical model has been developed to describe the mass transfer rate induced by buoyancy. Numerical simulations support the simple solution. A comparison is made with the regional gradient-driven flow model. It is shown that buoyancy-driven flow can noticeably increase the release rate.
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| 38. |
- Neretnieks, Ivars
(författare)
-
Radionuclide Transport in Channel Networks with Radial Diffusion in the Porous Rock Matrix
- 2023
-
Ingår i: Nuclear Technology. - : Informa UK Limited. - 0029-5450 .- 1943-7471. ; 209:4, s. 604-621
-
Tidskriftsartikel (refereegranskat)abstract
- Water flows in only a small fraction of the total area of the fractures in fractured rocks. The width of the “channels” is often in the range of centimeters to tens of centimeters. Nuclides can diffuse into and out of the porous rock matrix, which causes them to be significantly retarded compared to the water velocity. In discrete facture networks, diffusion is modeled to be linear and perpendicular to the fracture surface. From a narrow channel, the diffusion cloud would then be as wide as the channel. When the nuclide has propagated farther than the channel width, the diffusion will become essentially radial, which allows the nuclide flux to increase enormously. For the times of interest for a repository for high-level nuclide waste, this will increase nuclide flux into the matrix by tens to thousands of times, and consequently, the nuclide retardation in the flowing water. Radial diffusion was not invoked in the performance assessment of the Forsmark site, which in January 2022 was chosen by the government to locate Sweden’s high-level waste repository. It is shown, using data from this site, that the effect of radial diffusion from the narrow channels considerably increases the retardation of any escaping radionuclides, potentially allowing for the use of thinner copper canisters.
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| 39. |
- Nilsson, Karl-Fredrik, et al.
(författare)
-
A probabilistic methodology to determine failure probabilities and acceptance criteria for the KBS-3 inserts under ice-age load conditions
- 2008
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 163:1, s. 3-14
-
Tidskriftsartikel (refereegranskat)abstract
- The Swedish KBS-3 copper-cast iron canister for geological disposal of spent nuclear fuel is in an advanced stage. This paper deals with the cast iron insert that provides the mechanical strength of the canister and outlines an approach to assess the failure probabilities for manufactured canisters at large isostatic pressure (44 MPa) that could occur during future glaciations and first steps to derive acceptance criteria to ensure that failure probabilities are extremely small. The work includes a statistical test program using three inserts to determine the tensile, compression, and fracture properties. Specimens used for material characterization were also investigated by microstructural analysis to determine the microstructure and to classify and size defects. It was found that the material scatter and low ductility were caused by many defect types, but slag defects in the form of oxidation films were the most important ones. These data were then used to compute defect distributions for the probabilistic failure analysis of the KBS-3 canisters. A large number of finite element-analyses of canisters were performed at the maximum design load (44 MPa) covering distributions of material parameters and geometrical features of the canisters. The computed probabilities for fracture and plastic collapse were very low even for material data with low ductility. Two large-scale isostatic compression tests of KBS-3 mock-ups to verify safety margins are also described. The failure occurred at loads above 130 MPa in both cases, indicating a safety margin of at least a factor 3 against the maximum design load. As a result of the project, new acceptance criteria are being proposed for insert geometry and material properties, and the manufacturing process for inserts has been modified to ensure that these criteria are always fulfilled.
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| 40. |
- Painter, Scott L., et al.
(författare)
-
Time-domain random-walk algorithms for simulating radionuclide transport in fractured porous rock
- 2008
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 163:1, s. 129-136
-
Tidskriftsartikel (refereegranskat)abstract
- Time-domain random-walk (TDRW) algorithms are efficient methods for simulating solute transport along one-dimensional pathways. New extensions of the TDRW algorithm accommodate decay and ingrowth of radionuclides in a decay chain and time-dependent transport velocities. Tests using equilibrium sorption and matrix diffusion retention models demonstrate that the extended TDRW algorithm is accurate and computationally efficient. When combined with stochastic simulation of transport properties, the resulting algorithm, Particle on Random Streamline Segment (PORSS), also captures the effects of random spatial variations in transport velocities, including the effects of very broad velocity distributions. When used in combination with discrete fracture network simulations, the PORSS algorithm provides an accurate and practical method for simulating radionuclide transport at the geosphere scale without invoking the advection-dispersion equation.
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| 41. |
- Pasi, Anna-Elina, 1993, et al.
(författare)
-
Organic Telluride Formation from Paint Solvents Under Gamma Irradiation
- 2022
-
Ingår i: Nuclear Technology. - : Informa UK Limited. - 0029-5450 .- 1943-7471. ; 208:11, s. 1734-1744
-
Tidskriftsartikel (refereegranskat)abstract
- The interactions between tellurium and organic material during a nuclear reactor accident are critical to source term estimations because of the possible formation of volatile species. Reactions taking place in the containment sump are of interest since these can lead to re-volatilization and increase the fission product source term. This paper presents results from experiments investigating the interaction of tellurium dioxide with three paint solvents-texanol ester, methyl isobutyl ketone, and toluene-under containment sump conditions. The experiments were performed by irradiating a mixed solution of tellurium dioxide and paint solvents at a dose rate of 4 kGy/h up to 300 kGy. The resulting samples were analyzed for tellurium concentration and speciation. Tellurium(IV) was found to reduce to metallic tellurium under irradiation when paint solvents were present. More importantly, several volatile organic tellurides were identified in the irradiated samples, which suggests that tellurium can form volatile species in sump conditions when in contact with dissolved paint solvents. This paper provides novel evidence of organic telluride formation in the sump and raises further interest in tellurium chemistry during a severe nuclear reactor accident.
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| 42. |
- Pasi, Anna-Elina, 1993, et al.
(författare)
-
Tellurium Behavior in the Containment Sump: Dissolution, Redox, and Radiolysis Effects
- 2020
-
Ingår i: Nuclear Technology. - : Informa UK Limited. - 0029-5450 .- 1943-7471. ; , s. 1-11
-
Tidskriftsartikel (refereegranskat)abstract
- In the event of a severe nuclear accident, one major concern is the release of radioactive material into the environment causing potential exposure of the general public to radiation. Among the volatile radionuclides are a range of tellurium isotopes. Due to the radioactivity and the volatility of tellurium, it has to be taken into account when assessing the overall effects of an accident. The interest in tellurium is not limited only to its release but also to the fact that some tellurium isotopes decay to iodine, and thus affect the iodine release behavior. The release and transport behavior of tellurium has been investigated over the past decades, however, the aqueous chemistry of tellurium in the complex containment sump system is still unclear. This study presents the behavior of tellurium dioxide in simplified containment sump conditions in relation to dissolution, redox reactions, and interactions with water radiolysis products. The results indicate that radiolysis products have a significant effect on tellurium chemistry in both a reducing and oxidizing manner depending on the solution composition. The redox reactions also affect the solubility of tellurium. The results show that the current information used to assess tellurium source term needs to be reevaluated for both severe accident management and for code validation purposes.
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| 43. |
- Peltonen, Joanna, et al.
(författare)
-
Development of effective algorithm for coupled thermal-hydraulic-neutron- kinetics analysis of reactivity transient
- 2011
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 176:2, s. 195-210
-
Tidskriftsartikel (refereegranskat)abstract
- Analyses of nuclear reactor safety have increasingly required the coupling of full three-dimensional neutron-kinetics (NK) core models with system transient thermal-hydraulic (TH) codes. To produce results within a reasonable computing time, the coupled codes use different spatial descriptions of the reactor core. The TH code uses few, typically 5 to 20, TH channels that represent the core. The NK code uses the explicit node for each fuel assembly. Therefore, a spatial mapping of coarse grid TH and fine grid NK domain is necessary. However, improper mappings may result in the loss of valuable information, thus causing inaccurate prediction of safety parameters. The purpose of this investigation is to study the sensitivity of spatial coupling (channel refinement and spatial mapping) and develop recommendations for NK-TH mapping in the simulation of safety transients control rod drop, turbine trip, and feedwater transient combined with stability performance (minimum pump speed of recirculation pumps). The research methodology consists of a spatial coupling convergence study, as an increasing number of TH channels and different mapping schemes approach the reference case. The reference case consists of one TH channel per one fuel assembly. The comparison of results has been done under steady-state and transient conditions. The obtained results and conclusions are presented in this paper.
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| 44. |
- Preston, Markus, et al.
(författare)
-
Methodology for Multiparameter Evaluation of Barriers Against Proliferation of Minor Actinides
- 2024
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471.
-
Tidskriftsartikel (refereegranskat)abstract
- There exist elements apart from uranium and plutonium that could potentially be used to construct the core of a nuclear explosive device. These belong to the so-called minor actinides (MAs), which exist in nonnegligible amounts in spent nuclear fuel (SNF) and are in nearly all cases not covered by international safeguards. Future reprocessing of SNF could result in significant separation of these elements, potentially leading to new proliferation concerns. In this work, a methodology for a transparent assessment of the barriers against proliferation of MAs has been developed and applied to the case of neptunium, americium, and curium separated from spent fuel from pressurized water reactors. In this methodology, openly available data and Monte Carlo simulations have been used to assess the barriers posed by a number of parameters relevant to the production of a nuclear explosive device from SNF. The evaluation shows that the properties of neptunium present low barriers to proliferation and that it should be discussed within the context of future nonproliferation treaties and possibly be placed under international safeguards. The properties of americium and curium present higher barriers to proliferation, meaning that these elements require less focus in the nonproliferation context.
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| 45. |
|
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| 46. |
- Pusch, Roland
(författare)
-
Highly compacted sodium bentonite for isolating rock-deposited radioactive waste products
- 1979
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 45:2, s. 153-157
-
Tidskriftsartikel (refereegranskat)abstract
- Sodium saturated bentonite clay compacted to a high density is a very efficient isolation for preventing radiotoxic nuclides from deeply buried canisters with radioactive wastes from reaching the biosphere. The main function of the bentonite, which is applied in the form of blocks between the rock and the canisters in large boreholes, is to provide a practically impervious barrier. The bentonite blocks take up water and swell so that they completely fill the space between rock and canisters. The swelling potential, which is permanent, makes the bentonite self-sealing. This means that rock joints, which may be opened, are sealed by extruding bentonite.
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| 47. |
|
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| 48. |
- Rathore, Vikram, et al.
(författare)
-
First experimental demonstration of the use of a novel planar segmented HPGe detector for gamma emission tomography of mockup fuel rods
- 2023
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471.
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Tidskriftsartikel (refereegranskat)abstract
- Post-irradiation examination of nuclear fuel is routinely performed to characterize the important properties of the current and the future fuel. Gamma emission tomography is a proven non-invasive technique for this purpose. Among various measurement elements of the technique, a gamma-ray detector is an important element whose spectroscopic abilities and detection efficiency affect the overall results. Finding a combination of high detection efficiency and excellent energy resolution in a single detector is often a challenge. We have designed a novel planar segmented HPGe detector which offers simultaneous measurement in six lines of sight with excellent energy resolution. The simultaneous detection ability enables faster data acquisition in a tomographic measurement which may facilitate achieving higher spatial resolution. In this work, we have demonstrated the first use of the detector by performing a full tomographic measurement of mockup fuel rods. Two methods of detector data analysis were used to make spectra and the images (tomograms) were reconstructed using the filtered back projection algorithm. The reconstructed images validate the successful use of the detector for tomographic measurement. The use of the detector for real fuel measurement is being planned and will be performed in the near future.
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| 49. |
- Reisch, Frigyes
(författare)
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HIGH PRESSURE BOILING WATER REACTOR
- 2010
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 172:2, s. 101-107
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Tidskriftsartikel (refereegranskat)abstract
- Some 400 boiling water reactors (BWRs) and pressurized water reactors (PWRs) have been in operation for several decades. The presented concept, the high pressure boiling water reactor (HP-BWR), combines the best parts and omits the troublesome components of traditional BWRs and PWRs by taking into consideration the experiences gained during their operation. One of the major benefits of the HP-BWR is that safety is improved. The design utilizes gravity-operated control rods, and there is a large space for the cross-formed control rods between fuel boxes. The bottom of the reactor vessel is smooth and without penetrations. All the pipe connections to the reactor vessel are well above the top of the reactor core, and core spray is not needed. Additionally, internal circulation pumps are used. The HP-BWR concept is also environmentally friendly: Improved thermal efficiency is achieved by feeding the turbine with similar to 340 degrees C (15 MPa) steam instead of similar to 285 degrees C (7 MPa), and there is less warm water release to the recipient and less uranium consumption per produced kWh, resulting in the production of less waste. Finally, the HP-BWR is cost effective and simple, operating in direct cycle mode with no need for complicated steam generators. Moisture separators and steam dryers are placed inside the reactor vessel, and additional separators and dryers can be installed inside or outside the containment. Well-proved simple dry containment or wet containment can be used. In more than half a century, an extensive regulatory licensing experience has been built from traditional BWRs and PWRs. The HP-BWR is a developed, high-performance successor of those conventional designs. Therefore, it can be expected that licensing can be accomplished in a reasonable time. Several utilities are supporting manufacturers to study concepts for future reactors. It is likely that an application to one or more electrical power companies for financial support by a manufacturer to make a detailed feasibility study of the HP-BWR would be positively treated. This could be the next step to the implementation of the HP-BWR.
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| 50. |
- Soler, Josep M., et al.
(författare)
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Predictive and Inverse Modeling of a Radionuclide Diffusion Experiment in Crystalline Rock at ONKALO (Finland)
- 2023
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Ingår i: Nuclear Technology. - : Informa UK Limited. - 0029-5450 .- 1943-7471. ; 209:11, s. 1765-1784
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Tidskriftsartikel (refereegranskat)abstract
- The REPRO-TDE test was performed at a depth of about 400 m in the ONKALO underground research facility in Finland. Synthetic groundwater containing radionuclide tracers [tritiated water tracer (HTO), 36Cl, 22Na, 133Ba, and 134Cs] was circulated for about 4 years in a packed-off interval of the injection borehole. Tracer activities were additionally monitored in two observation boreholes. The test was the subject of a modeling exercise by the SKB GroundWater Flow and Transport of Solutes Task Force. Eleven teams participated in the exercise, using different model concepts and approaches. Predictive model calculations were based on laboratory-based information concerning porosities, diffusion coefficients, and sorption partition coefficients. After the experimental results were made available, the teams were able to revise their models to reproduce the observations. General conclusions from these back-analysis calculations include the need for reduced effective diffusion coefficients for 36Cl compared to those applicable to HTO (anion exclusion), the need to implement weaker sorption for 22Na compared to results from laboratory batch sorption experiments, and the observation of large differences between the theoretical initial concentrations for the strongly sorbing 133Ba and 134Cs, and the first measured values a few hours after tracer injection. Different teams applied different concepts, concerning mainly the implementation of isotropic versus anisotropic diffusion, or the possible existence of borehole disturbed zones around the different boreholes. The role of microstructure was also addressed in two of the models.
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