| 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. |
- 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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| 3. |
- 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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| 4. |
- 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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| 5. |
- 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
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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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| 6. |
- 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
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 146:1, s. 15-28
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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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| 7. |
- Liu, J. S., et al.
(författare)
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Effect of water radiolysis caused by dispersed radionuclides on oxidative dissolution of spent fuel in a final repository
- 2001
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 135:2, s. 154-161
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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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| 8. |
- Liu, J. S., et al.
(författare)
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Study of the consequences of secondary water radiolysis surrounding a defective canister
- 2003
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 142:3, s. 294-305
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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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| 9. |
- Liu, Longcheng, et al.
(författare)
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A coupled model for oxidative dissolution of spent fuel and transport of radionuclides from an initially defective canister
- 2001
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 135, s. 273-285
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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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| 10. |
- Liu, Longcheng, et al.
(författare)
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A reactive transport model for oxidative dissolution of spent fuel and release of nuclides within a defective canister
- 2002
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 137, s. 228-240
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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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