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Sample records for bwr fuel rods

  1. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  2. Optimization of fuel rod enrichment distribution for BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-09-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. the combinatorial optimization problem of grouping fuel rods into a given number of rod groups with the same enrichment, and the problem of determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by the linear combination C{sub 1}X + C{sub 2}X{sub G}, where X and X{sub G} stand, respectively, for control variables giving constraint to the local power peaking factor and the gadolinium rod power. C{sub 1} and C{sub 2} are user-definable weighting factors to accommodate design preferences. The algorithm for solving this combinatorial optimization problem starts by finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering. This latter problem is solved using the method of approximation programming. A practical application is shown for a contemporary 8 x 8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  3. Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

    International Nuclear Information System (INIS)

    The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

  4. Comparison of heat capacity and thermal time constant between BWR fuel and simulated heater rod

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    It is important to know the thermal characteristics of BWR fuel, i.e. heat capacity and thermal time constant, in order to evaluate the thermal hydraulics at BWR accidents and the events under thermal-hydraulic and neutronic coupling condition. Further, since the heater rod simulating BWR fuel is used in the tests for BWR accidents and for BWR thermal hydraulics coupled with neutronics, it is important to know the thermal characteristics of the heater rod. Therefore, the author investigated the thermal characteristics of BWR fuel and the heater rod by performing experiments and analyzing with J-TRAC code capable to analyze 2-dimensional heat conduction problem. The heat capacity per unit length of BWR fuel cp{rho}A (kJ/mK) was estimated to be 0.34 kJ/mK - 0.36 kJ/mK in 300 deg. C - 800 deg. C. The heat capacity of the heater rod was almost identical with each other regardless of the differences in rods and positions. It was higher with higher temperature. The heat capacity of the heater rod used in the test for BWR accidents was about 0.38 kJ/mK at 600 deg. C, which was about 9% higher than the average (0.35 kJ/mK) of BWR fuel. On the other hand, the heat capacity used in the test for BWR thermal hydraulics coupled with neutronics was about 0.42 kJ/mK at 600 deg. C, which was about 20% higher than the average of BWR fuel. Thermal time constant was affected by surface heat transfer coefficient, thermal diffusivity, and gap conductance. When the surface heat transfer coefficient is small, it controls the heat transfer and thermal time constant depends mainly on the surface heat transfer coefficient. When the surface heat transfer coefficient is large, the heat conduction controls the heat transfer and thermal time constant depends mainly on the thermal diffusivity. In the former case, one point heat transfer model is applicable and the thermal time constant is proportional to the inverse of the surface heat transfer coefficient. In this case, the thermal time

  5. Critical experiments for BWR fuel assemblies with cluster of gadolinia rods

    International Nuclear Information System (INIS)

    Gadolinia-bearing fuel rods are needed for high-burnup fuels. Strong neutron absorption of gadolinia makes an assembly heterogeneous from the viewpoint of reactor physics. The cluster of gadolinia-bearing fuel rods is useful for higher-burnup fuels than current fuels. Few critical experiments have been reported for fuel assemblies with the cluster of gadolinia-bearing fuel rods. We conducted critical experiments for BWR fuel assemblies with the cluster of gadolinia-bearing fuel rods in the Toshiba Nuclear Critical Assembly (NCA). Critical water level and power distribution were measured. Measurements were compared with analyses by a continuous-energy Monte Carlo code, MCNP, with the JENDL3.3 nuclear data library. (author)

  6. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    International Nuclear Information System (INIS)

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

  7. Entrainment and deposition modeling of liquid films with applications for BWR fuel rod dryout

    Science.gov (United States)

    Ratnayake, Ruwan Kumara

    While best estimate computer codes provide the licensing basis for nuclear power facilities, they also serve as analytical tools in overall plant and component design procedures. An ideal best estimate code would comprise of universally applicable mechanistic models for all its components. However, due to the limited understanding in these specific areas, many of the models and correlations used in these codes reflect high levels of empiricism. As a result, the use of such models is strictly limited to the range of parameters within which the experiments have been conducted. Disagreements between best estimate code predictions and experimental results are often explained by the mechanistic inadequacies of embedded models. Significant mismatches between calculated and experimental critical power values are common observations in the analyses of Boiling Water Reactors (BWR). Based on experimental observations and calculations, these mismatches are attributed to the additional entrainment and deposition caused by spacer grids in BWR fuel assemblies. In COBRA-TF (Coolant Boiling in Rod Arrays-Two Fluid); a state of the art industrial best estimate code, these disagreements are hypothesized to occur due the absence of an appropriate spacer grid model. In this thesis, development of a suitably detailed spacer grid model and integrating it to COBRA-TF is documented. The new spacer grid model is highly mechanistic so that the applicability of it is not seriously affected by geometric variations in different spacer grid designs. COBRA-TF (original version) simulations performed on single tube tests and BWR rod bundles with spacer grids showed that single tube predictions were more accurate than those of the rod bundles. This observation is understood to arise from the non-availability of a suitable spacer grid model in COBRA-TF. Air water entrainment experiments were conducted in a test section simulating two adjacent BWR sub channels to visualize the flow behavior at

  8. Optimization of fuel rod enrichment distribution to minimize rod power peaking throughout life within BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1997-01-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. a combinatorial optimization problem grouping fuel rods into a given number of rod groups with the same enrichment, and a problem determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by a linear combination: C{sub 1}X+C{sub 2}X{sub G}, where X and X{sub G} stand for a control variable to give the constraint respectively for a local power peaking factor and a gadolinium rod power, and C{sub 1} and C{sub 2} are user-definable weighting factor to accommodate the design preference. The algorithm of solving the combinatorial optimization problem starts with finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering, and then the latter problem is solved by using the method of approximation programming. The practical application of the present method is shown for a contemporary 8x8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  9. Hydrogen uptake of BWR fuel rods. Power history effects at long irradiation times

    International Nuclear Information System (INIS)

    AREVA LTP (Low Temperature Process) Zircaloy-2 cladding for Boiling Water Reactors (BWR) in both RXA (Recrystallized Annealed) and CWSR (Cold Worked Stress Relieved) metallurgical states, has an optimized microstructure with an optimum size of SPP (Secondary Phase Particles) that has reduced the nodular corrosion to a minimum while maintaining a good uniform corrosion performance with acceptable hydrogen pickup. Classically hydrogen uptake is described by the Hydrogen Pick-Up Fraction (HPUF), which is the ratio of the hydrogen generated by uniform oxidation that is eventually picked up by the metal to the total hydrogen generated by oxidation. In the past, the hydrogen uptake database showed a low HPUF with hydrogen concentration close to the saturation value of the metal at operating temperature and correspondingly little hydride formation. The hydrogen concentration was correlated with irradiation time via the HPUF (at an almost constant corrosion and hydrogen production rate). Recently, some significantly higher hydrogen concentration values (300 wppm and more) have been measured for medium and high burnup rods. This effect was also observed on four AREVA fuel rods from BWR (Boiling Water Reactors). This prompted a thorough analysis of the hydrogen pickup database as well as material and environmental factors influencing corrosion and hydrogen uptake. The most important outcome of the investigation was that a low power – low steam condition is associated with increased hydrogen pickup. The linear power is a proxy variable for low heat flux and low steam quality in the coolant, which were identified as important parameters for physical processes that could explain the enhanced hydrogen uptake in some cases. The paper will present the database of the enhanced hydrogen uptake measured in European power reactors and demonstrate the effect of power history on the uptake process. Power histories with high hydrogen uptake included extended low power periods later in

  10. Process development and fabrication for sphere-pac fuel rods. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

  11. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

    International Nuclear Information System (INIS)

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  12. BWR fuel performance

    International Nuclear Information System (INIS)

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  13. Experimental data report for test TS-5 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    This report presents experimental data for Test TS-5 which was the fifth test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in January, 1993. Test fuel rod used in the Test TS-5 was a short-sized BWR (7x7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79% and a burnup of 26GWd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The nominal energy deposition of 117±5cal/g·fuel (98±4cal/g·fuel in peak fuel enthalpy) was subjected to the test fuel rod and no fuel failure was observed in the test. The test fuel was pulse irradiated in a flow shroud which simulates fuel/water ratio in the commercial assembly. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

  14. BWR control rod patterns and fuel loading optimization using heuristic methods

    International Nuclear Information System (INIS)

    We show the results obtained with the OCOTH system to optimize the Fuel Reloads Design and Control Rod Patterns Design in a Boiling Water Reactor. Our system solves both problems in a coupled way. We used the 3-dimensional CM-PRESTO code to evaluate the solutions quality. The process has three stages. In the first step we obtain a Fuel Reload Design 'seed' using the Haling's principle. The followings steps are an iterative process between the Control Rod Patterns Designs and Fuel Reloads Design. Control Rod Patterns Design is proposed for the Fuel Reload Design 'seed' and then Control Rod Patterns Design is used to find a new Fuel Reload Design. Both processes are coupled in an iterative loop until a criterion stop is fulfilled. In the whole process, the genetic algorithms, neural networks and ant colony system optimization techniques were used. (authors)

  15. Comparison study of the thermal mechanical performance of fuel rods during BWR fuel preconditioning operations using the computer codes FUELSIM and FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Ortiz V, J.; Castillo D, R., E-mail: rafael.pantoja10@yahoo.com.m [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The safety of nuclear power plants requires monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behaviour under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. In the operation of a nuclear power reactor, pre-conditioning simulations are necessary to determine in advance limit values for the power that can be generated in a fuel rod during any power ramp, and mainly during reactor startup, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR is performed. This study includes two types of fuel rods: one from a fuel assembly design with array 8 x 8, and the other one from a 10 x 10 fuel assembly design, and a comparison of the thermal-mechanical performance between the two different rod designs is performed. The performance simulations were performed by the code FUELSIM, and compared against results previously obtained from similar simulation with the code FEMAXI-V. (Author)

  16. Fuel rod

    International Nuclear Information System (INIS)

    The present invention provide a fuel rod used in a BWR type reactor, preventing the occurrence of defects of weld portions and improving the operationability of test and assembling operation to improve the quality of weld portions. Namely, the fuel rod is formed by loading a plurality of fuel pellets in a cladding tube. The outer diameter of a groove portion of a tightly sealing end plug to be inserted and welded to the open end of the cladding tube is made substantially identical with the inner diameter of the cladding tube. A neck portion having a diameter smaller than the outer diameter of the groove portion is disposed between an end plug main body and the groove portion. As a result, since the outer diameter of the groove portion is substantially identical with the inner diameter of the cladding tube, the positioning is facilitated. Since the neck portion having a smaller diameter than the outer diameter of the groove portion is disposed in the groove portion, a gap is formed in the welded portion thereby enabling to facilitate the confirmation of weld sag for confirming integrity of the weld by a non-destructive test. (I.S.)

  17. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code

    International Nuclear Information System (INIS)

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  18. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  19. A comparison of crud phases appearing on some Swedish BWR fuel rods using Laser Raman Spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden)]|[Lulea Univ. of Technology (Sweden)

    2002-07-01

    Previous investigations showed that laser Raman spectroscopy (LRS) can be used as a phase specific analytical tool for radioactive fuel crud samples and also for details in the underlying layer of zirconium dioxide. It is relatively easy to record Raman spectra that discriminate between chemical phases for all crud oxides of interest. The method has therefore been recommended for crud investigations within the Swedish program. At ideal conditions the resolution is about 1 {mu}m, permitting detailed position determination of crud phases in the sample. Therefore LRS is a very good complement to X-ray diffraction (XRD). The methods for sample preparation and handling of radioactive crud samples for LRS turn out to be relatively simple. A detailed LRS study on fuel crud samples from Barsebaeck 2, Forsmark 2, Forsmark 3 and Ringhals 1 was performed in this work. All of those Swedish BWRs were operated at different conditions at the time of sampling. The chemistry regimes covered NWC, HWC and other variable conditions. Also different types of fuel, exposure times and sampling positions were selected. (authors)

  20. A comparison of crud phases appearing on some Swedish BWR fuel rods using Laser Raman Spectroscopy

    International Nuclear Information System (INIS)

    Previous investigations showed that laser Raman spectroscopy (LRS) can be used as a phase specific analytical tool for radioactive fuel crud samples and also for details in the underlying layer of zirconium dioxide. It is relatively easy to record Raman spectra that discriminate between chemical phases for all crud oxides of interest. The method has therefore been recommended for crud investigations within the Swedish program. At ideal conditions the resolution is about 1 μm, permitting detailed position determination of crud phases in the sample. Therefore LRS is a very good complement to X-ray diffraction (XRD). The methods for sample preparation and handling of radioactive crud samples for LRS turn out to be relatively simple. A detailed LRS study on fuel crud samples from Barsebaeck 2, Forsmark 2, Forsmark 3 and Ringhals 1 was performed in this work. All of those Swedish BWRs were operated at different conditions at the time of sampling. The chemistry regimes covered NWC, HWC and other variable conditions. Also different types of fuel, exposure times and sampling positions were selected. (authors)

  1. Transportation activities for BWR fuels at NFI

    International Nuclear Information System (INIS)

    Nuclear Fuel Industries, LTD. (NFI) supplies fuel assemblies for both PWR and BWR in Japan. We can also manage transportation of the fuel assemblies from our fabrication facilities to the nuclear power plants of Japanese utilities. For the transportation of fuel assembly, we designed and fabricated the transportation containers to meet the requirements of the IAEA regulations, and licensed in Japan. This paper introduced the recent activity and R and D of NFI concerning transportation of BWR fuel assembly. NT-XII transportation container was developed for fresh BWR fuel assemblies. NT-XII container consists of inner container and outer container. Two BWR fuel assemblies with up to 5 wt.-% 235-U enriched are enclosed in an inner container. In the concept of NT-XII container design, we made the best priority to transportation efficiency, as well as ensuring fuel integrity during transportation. NT-XII has been used since 2002 in Japan. Thanks to the lightening weight of containers, the number of containers to be loaded to one transportation truck was increased up to 9 containers (equivalent to 18 fuel assemblies) compared with former type container (NT-IV transportation container) which can be loaded up to 6 containers (equivalent to 12 fuel assemblies). In addition to the design of brand-new container, we promote the improvement of the packaging methods. In Japan, in order to reduce the damage to the fuel rod and fuel spacer while transporting, polyethylene sleeves which are called 'packing separators' are inserted in the rod-to-rod gap of fuel assembly. However, packing separators requires time and cost for the installation at fuel fabrication facility. In the same way, huge time and cost are needed for removal of packing separators at nuclear power plant. For the improvement of preparation efficiency before and after transportation, we investigated the influence of vibration to fuel integrity in case of transportation without packing separators. Based on the above

  2. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation

  3. Investigation of control rod worth and nuclear end of life of BWR control rods

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, Per

    2008-01-15

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of {sup 10}B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% {sup 10}B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in {sup 10}B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming.

  4. Analysis of the behavior of irradiated BWR fuel rod in storage dry conditions; Analisis del comportamiento de una barra combustible irradiada BWR en condiciones de almacenamiento en seco

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, A.; Montes, D.; Ruiz-Hervias, J.; Munoz-Reja, C.

    2014-07-01

    In order to complete previous studies of creep on PWR sheath material, developed a joint experimental program by CSN, ENRESA and ENUSA about BWR (Zircaloy-2) sheath material. This program consisted in creep tests and then on the material under creep, compression testing diametral obtaining the permissible displacement of the sheath to break. (Author)

  5. Experience and reliability of Framatome ANP's PWR and BWR fuel

    International Nuclear Information System (INIS)

    Based on three decades of fuel supply to 169 PWR and BWR plants on four continents, Framatome ANP has a very large database from operating experience feedback. The performance of Framatome PWR and BWR fuel is discussed for the period 1992-2001 with special emphasis on fuel failures, countermeasures and their effectiveness. While PWR fuel performance in most reactors has been good, the performance in some years did suffer from special circumstances that caused grid-to-rod fretting failures in few PWRs. After solving this problem, fuel of all types showed high reliability again. Especially the current PWR fuel products AFA 3G, HTP, Mark B and Mark BW showed a very good operating performance. Fuel reliability of Framatome ANP BWR fuel has been excellent over the last decade with average annual fuel rod failure rates under 1x10-5 since 1991. More than 40% of all BWR fuel failures in the 1992-2001 decade were caused by debris fretting. The debris problem has been remedied with the FUELGUARDTM lower tie plate, and by reactor operators' efforts to control the sources of debris. PCI, the main failure mechanism in former periods, affected only 10 rods. All of these rods had non-liner cladding. (author)

  6. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code; Evaluacion del desempeno termomecanico de barras de combustible de un reactor BWR durante una rampa de potencia utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R.

    2010-07-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  7. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  8. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    The BWR control rods made by ABB use boron carbide (B4C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B4C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  9. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type; Calculo de las razones de generacion de calor lineal que violen el limite termomecanico de deformacion plastica de la camisa en funcion del quemado de una barra combustible tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2003-07-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  10. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

    One of the methods to conduct transmutation of minor actinide is the use of BWR with thorium fuel. Thorium fuel has a specific behaviour of producing a little secondary minor actinides. Transmutation of minor actinide is done by loading it in the BWR with thorium fuel through two methods, namely close recycle and accumulation recycle. The calculation of minor actinide composition produced, weigh of minor actinide transmuted, and percentage of reminder transmutation was carried SRAC. The calculations were done to equivalent cell modeling from one fuel rod of BWR. The results show that minor actinide transmutation is more effective using thorium fuel than uranium fuel, through both close recycle and accumulation recycle. Minor actinide transmutation weight show that the same value for those recycle for 5th recycle. And most of all minor actinide produced from 5 unit BWR uranium fuel can transmuted in the 6th of close recycle. And, the minimal value of excess reactivity of the core is 12,15 % Δk/k, that is possible value for core operation

  11. Effect of thermal-hydraulic feedback on the BWR rod drop accident

    International Nuclear Information System (INIS)

    An important design-basis accident for boiling water reactors (BWR's) is the rod drop accident (RDA). This accident is defined to be a rapid reactor transient caused by an accidental drop (out of the core) of the highest-worth control rod at various conditions ranging from cold start-up to about 10% of rated power. For most BWR designs the highest worth rod is normally situated at the center of the core. Despite the fact that the chance of a RDA in extremely unlikely, the consequence of the RDA is of concern because of the potential for damage to fuel rods. Neglecting moderator feedback during the RDA is a poor assumption because energy is deposited in the fuel over a 3 to 4 second time period and hence there is time for heat to be conducted to the coolant. This may tend to ameliorate the accident considerably. Evaluation of the thermal-hydraulic feedback effect on the RDS in a BWR has been scarce in the literature. The object of this paper is to demonstrate the beneficial effect of thermal-hydraulic feedback in the RDA

  12. Fuel rod bowing

    International Nuclear Information System (INIS)

    The purpose of this investigation was to quantify the extent of fuel rod bowing in Westinghouse pressurized water reactors and to assess the effects of fuel rod bowing on plant safety and reliability. An empirical bow correlation was developed based on data from irradiated assemblies. Analyses conducted with these conservative empirical predictions show that: (1) generically identified DNBR margins are adequate to offset DNBR reductions due to rod bow, (2) the present design practice of increasing the highest calculated core peaking factor is sufficient to account for all deviations, including the effects of rod bow, and (3) fretting and corrosion of bowed rods are negligible. These conclusions indicate that fuel rod bowing results in no impact on plant safety or reliability

  13. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods; Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, G.L.

    1991-11-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degrees}C and whether the cladding of the stored spent fuel ever exceeds 350{degrees}C. Limiting the borehole to temperatures of 97{degrees}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degrees}C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 {times} 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degrees}C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350{degrees}C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft {times} 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40{degrees}C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation.

  14. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

  15. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  16. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  17. Protecting AREVA ATRIUM™ BWR fuel from debris fretting failure

    International Nuclear Information System (INIS)

    Historically, debris fretting has been the leading cause of fuel rod failure in BWR fuel assemblies, costing the industry millions of dollars in lost generation and negatively impacting the working area of plant site personnel. In this paper the focus will be on recent BWR fuel product innovation designed to eliminate debris related failures. Experience feedback from more than three decades of operation history with non-line-of-sight FUELGUARD™ lower tie plate debris filters will be presented. The development and relative effectiveness of successive generations of filtration technology will be discussed. It will be shown that modern, state of the art debris filters are an effective defense against debris fretting failure. Protective measures extend beyond inlet nozzle debris filters. The comprehensive debris resistance features built into AREVA’s newest fuel design, the ATRIUM™ 11, reduce the overall risk of debris entrapment as well as providing a degree of protection from debris that may fall down on the fuel assembly from above, e.g., during refueling operations. The positive recent experience in a debris sensitive plant will be discussed showing that the combination of advanced fuel technology and a robust foreign material exclusion program at the reactor site can eliminate the debris fretting failure mechanism. (author)

  18. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

  19. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Purpose: To enable a tight seal in fuel rods while keeping the sealing gas pressure at an exact predetermined pressure in fuel rods. Constitution: A vent aperture and a valve are provided to the upper end plug of a cladding tube. At first, the valve is opened to fill gas at a predetermined pressure in the fuel can. Then, a conical valve body is closely fitted to a valve seat by the rotation of a needle valve to eliminate the gap in the engaging thread portion and close the vent aperture. After conducting the reduced pressure test for the fuel rod in a water tank, welding joints are formed between the valve and the end plug through welding to completely seal the cladding tube. Since the welding is conducted after the can has been closed by the valve, the predetermined gas pressure can be maintained at an exact level with no efforts from welding heat and with effective gas leak prevention by the double sealing. (Kawakami, Y.)

  20. Fuel rod plugs

    International Nuclear Information System (INIS)

    Purpose: To prevent the formation of voids to the inside of welded portion in fuel rod plugs. Constitution: A fuel rod is tightly sealed by welding end plugs at both ends of a fuel can charged with nuclear fuel material. For the welding of the end plug, laser welding has now been employed with the reason of increasing the welding efficiency and reducing the welding heat distortion. However, if the end plug is laser-welded to the end of the fuel can in the conventional form, there is a problem that voids are liable to be formed near the deepest penetration in the welding portion. That is, gases evolved near the deepest penetration remains in a key-hole like welded metal portion to result in voids there. Accordingly, grooves capable of passing the laser beam key hole therethrough are disposed along the circumferential direction of the pipe at the end plug welded portion in the fuel can. In this way, since gases generating near the deepest penetration are discharged into the grooves, the key hole-like welded metal is completely filled and voids are not formed. (Kamimura, M.)

  1. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  2. Investigation of burnup credit implementation for BWR fuel

    International Nuclear Information System (INIS)

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)

  3. Application of Ultrasonic inspection Technique on Fuel Rod Seam Weld

    International Nuclear Information System (INIS)

    As of the end of March, 1996, 26 BWR power plants of which station capacity has reached up to 23,000 GW in total are in commercial operation in Japan, Japan Nuclear Fuel (JNF), a BWR fuel fabricator in Japan, has supplied fuels to those power plants for 25 years. This paper presents refinement of inspection technology applied to enhance completeness of fuel rod welding at JNF, which has cumulatively produced approximately 50,000 fuel bundles to the date in Japan. In this operation, TIG method has been employed for plug to tube welding of fuel rod, and X-ray radiography was formerly applied as nondestructive testing (NDT) means in order to verify weld integrity of every fuel rod. As there was limited capabilities of X-ray radiography such as shooting time and direction, and also inspection of fuel rod weld integrity is one of key characteristics of regulatory inspection according to the law, JNF has developed and applied more reliable and effective probe rotation type ultrasonic method. This paper presents refinement of inspection technology applied to enhance completeness of fuel rod welding at JNF

  4. ATRIUMTM Fuel - Continuous Upgrading for High Duty BWR Plants

    International Nuclear Information System (INIS)

    AREVA NP is a supplier of nuclear fuel assemblies and associated core components to Boiling Water Reactors worldwide, representing today more than 60 000 fuel assemblies. Since first delivered in 1992, ATRIUMTM10 fuel assemblies have now been supplied to a total of 32 BWR plants in the US, Europe, and Asia resulting in an operating experience over 20 000 fuel assemblies. Among them, the latest versions are ATRIUMTM 10XP and ATRIUMTM 10XM fuel assemblies which have been delivered to several utilities worldwide. During six years of operation experience reaching a maximum fuel assembly burnup of 66 MWd/kgU, no fuel failure of ATRIUMTM 10XP/XM occurred. Regular upgrading of the fuel assemblies' reliability and performance has been made possible thanks to AREVA NP's continuous improvement process and the 'Zero tolerance for failure' program. In this frame, the in-core behavior follow-up, manufacturing experience feedback and customer expectations are the bases for setting improvement management objectives. As an example, most fuel rod failures observed in the past years resulted from debris fretting and Pellet Cladding Interaction (PCI) generally caused by Missing Pellet Surface. To address these issues, the development of the Improved FUELGUARDTM debris filter was initiated and completed while implementation of chamfered pellets and Cr doped fuel will address PCI aspects. In the case of fuel channel bow issue, efforts to ensure dimensional stability at high burnup levels and under challenging corrosion environments have been done resulting in material recommendations and process developments. All the described solutions will strongly support the INPO goal of 'Zero fuel failures by 2010'. In a longer perspective, the significant trend in nuclear fuel operation is to increase further the discharge burnup and/or to increase the reactor power output. In the majority of nuclear power plants worldwide, strong efforts in power up-rating were made and are still ongoing. Most

  5. BWR fuel cycle optimization using neural networks

    International Nuclear Information System (INIS)

    Highlights: → OCONN a new system to optimize all nuclear fuel management steps in a coupled way. → OCON is based on an artificial recurrent neural network to find the best combination of partial solutions to each fuel management step. → OCONN works with a fuel lattices' stock, a fuel reloads' stock and a control rod patterns' stock, previously obtained with different heuristic techniques. → Results show OCONN is able to find good combinations according the global objective function. - Abstract: In nuclear fuel management activities for BWRs, four combinatorial optimization problems are solved: fuel lattice design, axial fuel bundle design, fuel reload design and control rod patterns design. Traditionally, these problems have been solved in separated ways due to their complexity and the required computational resources. In the specialized literature there are some attempts to solve fuel reloads and control rod patterns design or fuel lattice and axial fuel bundle design in a coupled way. In this paper, the system OCONN to solve all of these problems in a coupled way is shown. This system is based on an artificial recurrent neural network to find the best combination of partial solutions to each problem, in order to maximize a global objective function. The new system works with a fuel lattices' stock, a fuel reloads' stock and a control rod patterns' stock, previously obtained with different heuristic techniques. The system was tested to design an equilibrium cycle with a cycle length of 18 months. Results show that the new system is able to find good combinations. Cycle length is reached and safety parameters are fulfilled.

  6. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  7. BWR Fuel Lattice Design Using an Ant Colony Model

    International Nuclear Information System (INIS)

    This paper deals with one of the steps of the nuclear fuel design: the radial fuel lattice design. It can be seen as a combinatorial optimization problem for determining the optimal 2D fuel rods enrichment and gadolinia distribution. In order to solve this optimization problem, the ant colony system technique is proposed. The main idea of the ant colony approach consists of emulating the real ant colony behaviour in their searching for minimum paths between two given points, usually between the nest and a food source. In this case, the environment where the artificial ants move is the space defined by the discrete possible values of Gd2O3 contents, the U235 enrichment, and the valid locations inside the 10x10 BWR fuel lattice array. In order to assess any candidate fuel lattice in the optimization process, the HELIOS neutronic transport code is used. The results obtained in the application of the implemented model show that the proposed technique is a powerful tool to tackle this step of the fuel design. (authors)

  8. BWR Fuel Lattice Design Using an Ant Colony Model

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose L.; Ortiz, Juan J. [Instituto Nacional de Investigaciones Nucleares, Depto. de Sistemas Nucleares, Carretera Mexico Toluca S/N. La Marquesa Ocoyoacac. 52750, Estado de Mexico (Mexico); Francois, Juan L.; Martin-del-Campo, Cecilia [Depto. de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico Paseo Cuauhnahuac 8532. Jiutepec, Mor. 62550 (Mexico)

    2008-07-01

    This paper deals with one of the steps of the nuclear fuel design: the radial fuel lattice design. It can be seen as a combinatorial optimization problem for determining the optimal 2D fuel rods enrichment and gadolinia distribution. In order to solve this optimization problem, the ant colony system technique is proposed. The main idea of the ant colony approach consists of emulating the real ant colony behaviour in their searching for minimum paths between two given points, usually between the nest and a food source. In this case, the environment where the artificial ants move is the space defined by the discrete possible values of Gd{sub 2}O{sub 3} contents, the U{sup 235} enrichment, and the valid locations inside the 10x10 BWR fuel lattice array. In order to assess any candidate fuel lattice in the optimization process, the HELIOS neutronic transport code is used. The results obtained in the application of the implemented model show that the proposed technique is a powerful tool to tackle this step of the fuel design. (authors)

  9. Sphere-pac versus pellet UO2 fuel in de Dodewaard BWR

    International Nuclear Information System (INIS)

    Comparative testing of UO2 sphere-pac and pellet fuel rods under LWR conditions has been jointly performed by the Netherlands Utilities Research Centre (KEMA) in Arnhem, the Netherlands Energy Research Foundation (ECN) at Petten and the Netherlands Joint Nuclear Power Utility (GKN) at Dodewaard. This final report summarizes the highlights of this 1968-1988 program with strong emphasis on the fuel rods irradiated in the Dodewaard BWR. The conclusion reached is that under normal LWR conditions sphere-pac UO2 in LWR fuel rods offers better resistance against stress corrosion cracking of the cladding, but that under fast, single step, power ramping conditions pellet UO2 in LWR fuel rods has a better resistance against hoop stress failure of the cladding. 128 figs., 36 refs., 19 tabs

  10. Proving test on thermal-hydraulic performance of BWR fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation (NUPEC) has conducted a proving test for thermal-hydraulic performance of BWR fuel (high-burnup 8 x 8, 9 x 9) assemblies entrusted by the Ministry of Economy, Trade and Industry (NUPEC-TH-B Project). The high-burnup 8 x 8 fuel (average fuel assembly discharge burnup: about 39.5 GWd/t), has been utilized from 1991. And the 9 x 9 fuel (average fuel assembly discharge burnup: about 45 GWd/t), has started to be used since 1999. There are two types (A-type and B-type) of fuel design in 9 x 9 fuel assembly. Using an electrically heated test assembly which simulated a BWR fuel bundle on full scale, flow induced vibration, pressure drop, critical power under steady state condition and post-boiling transition (post-BT) tests were carried out in an out-of pile test facility that can simulate the high pressure and high temperature conditions of BWRs. This paper completed the results of 9 x 9 fuel combined with the previously reported results of high-burnup 8 x 8 fuel. As a result of NUPEC-TH-B Project, the validity of the current BWR thermal-hydraulic design method was confirmed and the reliability of BWR thermo-hydraulic fuel performance was demonstrated. Based on the test data, a new correlation of the estimation of fuel rod vibration amplitude, new post-BT heat transfer and rewet correlations for the estimation of fuel rod surface temperature were developed. (author)

  11. Control Rod Pattern Planning of a BWR using Enhanced Nelder-Mead Method

    International Nuclear Information System (INIS)

    We propose a new optimization algorithm for the short-term planning of control rod patterns in an operating BWR. This algorithm is based on the enhanced Nelder-Mead simplex method in which convergence ability is improved for constrained problems in several ways. The main characteristic of this approach is it uses continuous values for the axial positions of control rods. Through calculations in an actual BWR plant, we showed that the new algorithm is effective for automation of short-term planning and reduction of the engineer's workload. (authors)

  12. A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

    Science.gov (United States)

    Chandraker, Dinesh Kumar; Vijayan, Pallipattu Krishnan; Sinha, Ratan Kumar; Aritomi, Masanori

    The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were

  13. LOFT nuclear fuel rod behavior

    International Nuclear Information System (INIS)

    An overview of the calculational models used to predict fuel rod response for Loss-of-Fluid Test (LOFT) data from the first LOFT nuclear test is presented and discussed and a comparison of predictions with experimental data is made

  14. Simulation of leaking fuel rods

    International Nuclear Information System (INIS)

    The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of 123I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123I release from failed fuel rods during transients

  15. On the domestic fuel channel for BWR

    International Nuclear Information System (INIS)

    Kobe Steel Ltd. started the domestic manufacture of fuel channel boxes for BWRs in 1967, and entered the actual production stage four years after that. Since 1976, the mass production system was adopted with the increase of the demand. The requirements about the surface contamination and the dimensional accuracy over whole length are very strict in the fuel channel boxes, moreover, special consideration must be given so as to prevent the deformation in use. The unique working methods such as electron beam welding, high temperature press forming and so on are employed in Kobe Steel Ltd. to satisfy such strict requirements, therefore the quality of the produced fuel channel boxes is superior to imported ones. At present, the fuel channel boxes domestically made by Kobe Steel Ltd. are used for almost all BWRs in Japan. The functions of fuel channel boxes are to flow boiling coolant uniformly upward, to guide control rods, and to increase the rigidity of fuel assembly. The fuel channel boxes are the square tubes of zircaloy 4 of 134.06 mm inside width, 2.03 mm thickness, and 4118 or 4239 mm length. The progress of the development and the features of the fuel channel boxes and the manufacturing processes are described. Zircaloy plates are formed into channels, and two channels are electron beam-welded after the edge preparation, to make a box. Ultrasonic examination and stress relief treatment are applied, and clips and spacers are welded. (Kako, I.)

  16. Huitzoctli: A system to design Control Rod Pattern for BWR's using a hybrid method

    International Nuclear Information System (INIS)

    Highlights: → The system was developed to design Control Rod Patterns for Boiling Water Reactors. → The critical reactor core and the thermal limits were fulfilled in all tested cases. → The Fuel Loading Pattern remains without changes during the iterative process. → The system uses the heuristics techniques: Scatter Search and Tabu Search. → The effective multiplication factor keff at the EOC was improved in all tested cases. - Abstract: Huitzoctli system was developed to design Control Rod Patterns for Boiling Water Reactors (BWR). The main idea is to obtain a Control Rod Pattern under the following considerations: (a) the critical reactor core state is satisfied, (b) the axial power distribution must be adjusted to a target axial power distribution proposal, and (c) the maximum Fraction of Critical Power Ratio (MFLCPR), the maximum Fraction of Linear Power Density (FLPD) and the maximum Fraction of Average Planar Power Density (MPGR) must be fulfilled. Those parameters were obtained using the 3D CM-PRESTO code. In order to decrease the problem complexity, Control Cell Core load strategy was implemented; in the same way, intermediate axial positions and core eighth symmetry were took into account. In this work, the cycle length was divided in 12 burnup steps. The Fuel Loading Pattern is an input data and it remains without changes during the iterative process. The Huitzoctli system was developed to use the combinatorial heuristics techniques Scatter Search and Tabu Search. The first one was used as a global search method and the second one as a local search method. The Control Rod Patterns obtained with the Huitzoctli system were compared to other Control Rod Patterns designs obtained with other optimization techniques, under the same operating conditions. The results show a good performance of the system. In all cases the thermal limits were satisfied, and the axial power distribution was adjusted to the target axial power distribution almost

  17. Refabricated and instrumented fuel rods

    International Nuclear Information System (INIS)

    Nuclear Fuel for power reactors capabilities evaluation is strongly based on the intimate knowledge of its behaviour under irradiation. This knowledge can be acquired from refabricated and instrumented fuel rods irradiated at different levels in commercial reactors. This paper presents the development and qualification of a new technique called RECTO related to a double-instrumented rod re-fabrication process developed by CEA/LECA hot laboratory facility at CADARACHE. The technique development includes manufacturing of the properly dimensioned cavity in the fuel pellet stack to house the thermocouple and the use of a newly designed pressure transducer. An analytic irradiation of such a double-instrumented fuel rod will be performed in OSIRIS test reactor starting October 2004. (Author)

  18. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  19. Droplet entrainment and deposition rate models for determination of boiling transition in BWR fuel assembly

    International Nuclear Information System (INIS)

    Droplet entrainment and deposition rates are of vital importance for mechanistic determination of critical power and location of boiling transition in a BWR fuel assembly. Data from high-pressure, high-temperature steam-water adiabatic experiments conducted in very tall test sections are used to develop a combination of equilibrium entrainment-deposition rate. Application of this combination to the heated tests conducted in a shorter test section of typical height of a BWR fuel assembly shows that correct split of total liquid in form of the film and droplets at the onset of annular-mist flow regime is also important to obtain good prediction of film flow rates/entrainment fraction. The improved model is then applied to simulate critical power tests in annulus and rod bundles. (author)

  20. Liquid films and droplet deposition in a BWR fuel element

    International Nuclear Information System (INIS)

    In the upper part of boiling water reactors (BWR) the flow regime is dominated by a steam-water droplet flow with liquid films on the nuclear fuel rod, the so called (wispy) annular flow regime. The film thickness and liquid flow rate distribution around the fuel rod play an important role especially in regard to so called dryout, which is the main phenomenon limiting the thermal power of a fuel assembly. The deposition of droplets in the liquid film is important, because this process sustains the liquid film and delays dryout. Functional spacers with different vane shapes have been used in recent decades to enhance droplet deposition and thus create more favorable conditions for heat removal. In this thesis the behavior of liquid films and droplet deposition in the annular flow regime in BWR bundles is addressed by experiments in an adiabatic flow at nearly ambient pressure. The experimental setup consists of a vertical channel with the cross-section resembling a pair of neighboring subchannels of a fuel rod bundle. Within this double subchannel an annular flow is established with a gas-water mixture. The impact of functional spacers on the annular flow behavior is studied closely. Parameter variations comprise gas and liquid flow rates, gas density and spacer shape. The setup is instrumented with a newly developed liquid film sensor that measures the electrical conductance between electrodes flush to the wall with high temporal and spatial resolution. Advanced post-processing methods are used to investigate the dynamic behavior of liquid films and droplet deposition. The topic is also assessed numerically by means of single-phase Reynolds-Averaged-Navier-Stokes CFD simulations of the flow in the gas core. For this the commercial code STAR-CCM+ is used coupled with additional models for the liquid film distribution and droplet motion. The results of the experiments show that the liquid film is quite evenly distributed around the circumference of the fuel rods. The

  1. Fuel rod failure due to marked diametral expansion and fuel rod collapse occurred in the HBWR power ramp experiment

    International Nuclear Information System (INIS)

    In the power ramp experiment with the BWR type light water loop at the HBWR, the two pre-irradiated fuel rods caused an unexpected pellet-cladding interaction (PCI). One occurred in the fuel rod with small gap of 0.10 mm, which was pre-irradiated up to the burn-up of 14 MWd/kgU. At high power, the diameter of the rod was increased markedly without accompanying significant axial elongation. The other occurred in the rod with a large gap of 0.23 mm, which was pre-irradiated up to the burn-up of 8 MWd/kgU. The diameter of the rod collapsed during a diameter measurement at the maximum power level. The causes of those were investigated in the present study by evaluating in-core data obtained from equipped instruments in the experiment. It was revealed from the investigation that these behaviours were attributed to the local reduction of the coolant flow occurred in the region of a transformer in the ramp rig. The fuel cladding material is seemed to become softened due to temperature increase caused by the local reduction of the coolant flow, and collapsed by the coolant pressure, either locally or wholly depending on the rod diametral gap existed. (author)

  2. Aging and service wear of control rod drive mechanisms for BWR nuclear plants

    International Nuclear Information System (INIS)

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''managing'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to normal wear and aging are the Graphiter seals. The predominant causes of aging for these seals are mechanical wear and thermally induced embrittlement More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are valve seals, discs, seats, stems, packing, and diaphragms. Since CRDM changeout and rebuilding is one of the highest dose, most physically challenging, and complicated maintenance activities routinely accomplished by BWR utilities, this report also highlights recent innovations in CRDM handling equipment and rebuilding tools that have resulted in significant dose reductions to the maintenance crews using them

  3. Leaf spring puller for nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Fogg, J.L.

    1981-11-03

    A fuel rod puller in the form of a collet for pulling fuel rods from a storage area into grids of a nuclear reactor fuel assembly. The rod puller moves longitudinally through the grids to a storage area where projections on the end of leaf springs grasp onto an end plug in a fuel rod. Drive apparatus then pulls the rod puller and connected fuel rod from the storage area into the fuel assembly grids. The rod puller includes an outer tube having leaf springs on one end thereof in one modification, mounted within the outer tube is a movable plunger which acts to urge the leaf springs outwardly to a position to permit passing or with the end of a end plug. Upon withdrawal of the plunger, the leaf springs move into a groove formed in the end of a fuel rod end plug, and the fuel rod subsequently is pulled into the fuel assembly grids. In another modification, the leaf springs on the outer rod are biased in an outward direction and a longitudinally movable tube on the outer rod is moved in a direction to contract the leaf springs into a position where the projections thereof engage the groove formed in a fuel rod end plug.

  4. Vibrational characteristics and wear of fuel rods

    International Nuclear Information System (INIS)

    Fuel rod wear, due to vibration, is a continuing concern in the design of liquid-cooled reactors. In my report, the methodology and models that are used to predict fuel rod vibrational response and vibratory wear, in a light water reactor environment, are discussed. This methodology is being followed at present in the design of Westinghouse Nuclear Fuel. Fuel rod vibrations are expressed as the normal bending modes, and sources of rod vibration are examined with special emphasis on flow-induced mechanisms in the stable flow region. In a typical Westinghouse PWR fuel assembly design, each fuel rod is supported at multiple locations along the rod axis by a square-shaped 'grid cell'. For a fuel rod /grid support system, the development of small oscillatory motions, due to fluid flow at the rod/grid interface, results in material wear. A theoretical wear mode is developed using the Archard Theory of Adhesive Wear as the basis. Without question certainty, fretting wear becomes a serious problem if it progresses to the stage where the fuel cladding is penetrated and fuel is exposed to the coolant. Westinghouse fuel is designed to minimize fretting wear by limiting the relative motion between the fuel rod and its supports. The wear producing motion between the fuel rod and its supports occurs when the vibration amplitude exceeds the slippage threshold amplitude

  5. Fuel rod pressure in nuclear power reactors: Statistical evaluation of the fuel rod internal pressure in LWRs with application to lift-off probability

    International Nuclear Information System (INIS)

    In this thesis, a methodology for quantifying the risk of exceeding the Lift-off limit in nuclear light water power reactors is outlined. Due to fission gas release, the pressure in the gap between the fuel pellets and the cladding increases with burnup of the fuel. An increase in the fuel-clad gap due to clad creep would be expected to result in positive feedback, in the form of higher fuel temperatures, leading to more fission gas release, higher rod pressure, etc, until the cladding breaks. An increase in the fuel-clad gap that leads to this positive feedback is a phenomenon called Lift-off and is a limitation that must be considered in the fuel core management. Lift-off is a consequence of very high internal fuel rod pressure. The internal fuel rod pressure is therefore used as a Lift-off indicator. The internal fuel rod pressure is closely connected to the fission gas release into the fuel rod plenum and is thus used to increase the database. It is concluded that the dominating error source in the prediction of the pressure in Boiling Water Reactors (BWR), is the power history. There is a bias in the fuel pressure prediction that is dependent on the fuel rod position in the fuel assembly for BWRs. A methodology to quantify the risk of the fuel rod internal pressure exceeding a certain limit is developed; the risk is dependent of the pressure prediction and the fuel rod position. The methodology is based on statistical treatment of the discrepancies between predicted and measured fuel rod internal pressures. Finally, a methodology to estimate the Lift-off probability of the whole core is outlined

  6. Fuel rod pressure in nuclear power reactors: Statistical evaluation of the fuel rod internal pressure in LWRs with application to lift-off probability

    Energy Technology Data Exchange (ETDEWEB)

    Jelinek, Tomas

    2001-02-01

    In this thesis, a methodology for quantifying the risk of exceeding the Lift-off limit in nuclear light water power reactors is outlined. Due to fission gas release, the pressure in the gap between the fuel pellets and the cladding increases with burnup of the fuel. An increase in the fuel-clad gap due to clad creep would be expected to result in positive feedback, in the form of higher fuel temperatures, leading to more fission gas release, higher rod pressure, etc, until the cladding breaks. An increase in the fuel-clad gap that leads to this positive feedback is a phenomenon called Lift-off and is a limitation that must be considered in the fuel core management. Lift-off is a consequence of very high internal fuel rod pressure. The internal fuel rod pressure is therefore used as a Lift-off indicator. The internal fuel rod pressure is closely connected to the fission gas release into the fuel rod plenum and is thus used to increase the database. It is concluded that the dominating error source in the prediction of the pressure in Boiling Water Reactors (BWR), is the power history. There is a bias in the fuel pressure prediction that is dependent on the fuel rod position in the fuel assembly for BWRs. A methodology to quantify the risk of the fuel rod internal pressure exceeding a certain limit is developed; the risk is dependent of the pressure prediction and the fuel rod position. The methodology is based on statistical treatment of the discrepancies between predicted and measured fuel rod internal pressures. Finally, a methodology to estimate the Lift-off probability of the whole core is outlined.

  7. Operation tests of parameters and new methods of the fuel rod production

    International Nuclear Information System (INIS)

    From 1974 to 1977 510 pathfinder fuel rods were inserted into the pressurized water reactor of Obrigheim (KWO) as part of several fuel reloads. They were irradiated for one to three reactor periods without indications of defects. The maximum burnup reached is about 36000 MWd/tU. In addition 80 fuel rods of the first core of the BWR power plant Isar (KKI, new fuel assembly type 8x8) were exactly precharacterized in 1977, to provide a good basis for post-irradiation examinations. (orig.) 891 HP/orig. 892 MKO

  8. Axial diagnostic system of finished rods BWR type

    International Nuclear Information System (INIS)

    This system is employed as a final non destructive diagnostic system to verify the adequate distribution of the different enrichment through the can of nuclear fuel. The system is framed of traction mechanisms, a personal computer, a counting card and another card for the pass motor control, the nuclear electronics and the control program. The performance is based on the gamma radiation counting of the natural decay of uranium 235, this radiation is processed by the nuclear instrumentation for delivering a pulse by each gamma detected. (Author)

  9. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  10. Fuel Thermal Expansion (FTHEXP). [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Reymann, G. A.

    1978-07-01

    A model is presented which deals with dimensional changes in LWR fuel pellets caused by changes in temperature. It is capable of dealing with any combination of UO/sub 2/ and PuO/sub 2/ in solid, liquid or mixed phase states, and includes expansion due to the solid-liquid phase change. The function FTHEXP models fuel thermal expansion as a function of temperature, fraction of PuO/sub 2/, and the fraction of fuel which is molten.

  11. Development of neural network simulating power distribution of a BWR fuel bundle

    International Nuclear Information System (INIS)

    A neural network model is developed to simulate the precise nuclear physics analysis program code for quick scoping survey calculations. The relation between enrichment and local power distribution of BWR fuel bundles was learned using two layers neural network (ENET). A new model is to introduce burnable neutron absorber (Gadolinia), added to several fuel rods to decrease initial reactivity of fresh bundle. The 2nd stages three layers neural network (GNET) is added on the 1st stage network ENET. GNET studies the local distribution difference caused by Gadolinia. Using this method, it becomes possible to survey of the gradients of sigmoid functions and back propagation constants with reasonable time. Using 99 learning patterns of zero burnup, good error convergence curve is obtained after many trials. This neural network model is able to simulate no learned cases fairly as well as the learned cases. Computer time of this neural network model is about 100 times faster than a precise analysis model. (author)

  12. Study of transient rod extraction failure without RBM in a BWR

    International Nuclear Information System (INIS)

    The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

  13. Local Fuel Rod Crud Prediction Tool Applications

    Energy Technology Data Exchange (ETDEWEB)

    Krammen, Michael A.; Karoutas, Zeses E.; Wang, Guoqiang; Young, Michael Y

    2009-06-15

    A code system with attendant methods has been developed for modeling local fuel rod crud. This tool is used to perform the Crud Induced Localized Corrosion (CILC) risk assessment recommended by the EPRI crud and corrosion guidelines, which were developed in response to the INPO zero fuel failures by 2010 initiatives. The methodology is in production use. This paper will describe the range of problems the methodology has already been applied to and the especial pertinence to low duty fuel applications. The methodology begins with Computational Fluid Dynamics (CFD) computations over a fuel assembly grid span. The CFD results provide detailed relative variations in local heat transfer coefficient over the grid span. These very local relative variations are used to determine very local thermal hydraulic conditions over the entire axial length of every fuel rod in a reactor core over the life of the rod in reactor. The expansion using the local relative variations is currently accomplished with the HIDUTYDRV code. The very local thermal hydraulic conditions are combined with reactor coolant crud concentrations derived from EPRI BOA analysis as input to models for predicting very local fuel rod crud deposition. The reactor coolant crud concentrations are determined over each reactor cycle by reactor system wide crud mass balance calculations. The reactor coolant crud concentrations are used to calculate local crud thickness using mass transfer models which are a function of the local thermal conditions. The advanced crud deposition models also include models for calculating local crud dryout. Local crud deposition and crud dryout are strongly dependent on very local boiling or steaming, which are predicted through the translation of the CFD results. The local crud thickness and degree of local crud dryout are key factors in determining the margin or risk for local fuel rod cladding crud induced fuel failure. The development and first application of these methods was in

  14. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M. (Argonne National Lab., IL (USA))

    1989-09-01

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 {mu}m in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307{degree}C rather than the normal 288{degree}C, a relatively thick (50 to 70 {mu}m) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs.

  15. Radial power density distribution of MOX fuel rods in the IFA-651

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Ho; Koo, Yang Hyun; Joo, Hyung Kook; Cheon, Jin Sik; Oh, Je Yong; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    Two MOX fuel rods, which were fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with Korea Atomic Energy Research Institute, have been irradiated in the HBWR from June, 2000 in the framework of OECD-HRP together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is basic in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR{sub H}BWR that calculates radial power density distribution for three MOX fuel rods has been developed based on neutron physics results and DEPRESS program. The developed subroutine FACTOR{sub H}BWR gives good agreement with the physics calculation except slight under-prediction at the outer part of the pellet above the burnup of 20 MWd/kgHM. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. 24 figs., 4 tabs. (Author)

  16. COBRA-SFS [Spent-Fuel Storage] thermal-hydraulic analyses of the CASTOR-1C and REA 2023 BWR storage casks containing consolidated spent fuel

    International Nuclear Information System (INIS)

    Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates

  17. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    International Nuclear Information System (INIS)

    A denatured (U-233/Th)O2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O2-fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O2-fueled BWR should perform similar to a UO2-fueled BWR under all operating conditions. A (Pu/Th)O2-fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO2-fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  18. Probabilistic assessment for nuclear fuel rods behavior

    International Nuclear Information System (INIS)

    BACO is a code for the simulation of the thermo-mechanical and fission gas behavior of a cylindrical fuel rod under operation conditions. Input parameters and, therefore, output ones may include statistical dispersion. In this paper, experimental CANDU fuel rods irradiated at the NRX reactor together with experimental MOX fuel rods and the IAEA'CRP FUMEX cases are used in order to determine the sensitivity of BACO code predictions. We analyze the CARA and CAREM fuel rods relation between predicted performance and statistical dispersion in order of enhanced their original designs. These exercises show the sensitivity of the predictions concerning such parameters and the extended features of the BACO code for a probability study. (author)

  19. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

    International Nuclear Information System (INIS)

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  20. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  1. Fuel rod welding (LWBR development program)

    International Nuclear Information System (INIS)

    Procedures were developed to weld both ends of approximately 25,000 fuel rods for the Light Water Breeder Reactor (LWBR) core. The rods were welded using the gas tungsten arc (GTA) method in high-purity helium at 1 atmosphere. Welding parameters, including weld current, arc gap, and speed of rotation, were established to control the size of the weld. Electrode and chill positioning with respect to the endclosure/tube joint controlled the location of the weld. Weld quality of the fuel rods was ensured by 100-percent nondestructive testing by ultrasonic and radiographic inspection and the destructive evaluation of process control samples in each weld lot

  2. EB welding and quality control of nuclear reactor fuel rods at ASEA-ATOM

    International Nuclear Information System (INIS)

    Fourteen years ago ASEA-ATOM chose EB welding for fuel rod plug/tube welds. This choice was made on the basis of 7 years of experience of EB-welding of fuel rods in a pilot plant. The specific reasons were the high quality and the high process yield, which are made possible by the great degree of controlability and reproducibility of this process and because the welds are suitable for QC inspection by an inline ultrasonic method which we developed at the same time. To date ASEA-ATOM has manufactured approximately 600,000 fuel rods with 1,200,000 EB-welds. The results have met expections as regards quality, process yield and service in BWR and PWR reactors. Descriptions are given of the automatic Sciaky EB welding machines, of the ultrasonic inspection equipment and of their process qualification. Some comments are made on quality and process yield

  3. Nuclear reactor fuel assembly with fuel rod removal means

    International Nuclear Information System (INIS)

    A fuel assembly is described for a nuclear reactor. The assembly has a bottom nozzle, at least one longitudinally extending control rod guide thimble attached to and projecting upwardly from the bottom nozzle and transverse grids spaced along the thimble. An organized array of elongated fuel rods are transversely spaced and supported by the grids and axially captured between the bottom nozzle and a top nozzle. The assembly comprises: (a) a transversely extending adapter plate formed by an arrangement of integral cross-laced ligaments defining a plurality of coolant flow openings; (b) means for mounting the adapter plate on an upper end portion of the thimble and spaced axially above and disposed transversely over the upper ends of all of the fuel rods present in the fuel assembly such that ones of the ligaments overlie corresponding ones of the fuel rods so as to prevent the fuel rods from moving upwardly through the coolant flow openings; and (c) removable plug means confined within the adapter plate and positioned over and spaced axially above selected ones of the fuel rods in providing access to at least one fuel rod for removal thereof upwardly through the axially spaced adapter plate without removing the top nozzle from the fuel assembly

  4. Three dimensional considerations in thermal-hydraulics of helical cruciform fuel rods for LWR power uprates

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush, E-mail: kshirvan@mit.edu; Kazimi, Mujid S.

    2014-04-01

    Highlights: • We benchmarked the 4 × 4 helical cruciform fuel (HCF) bundle pressure drop experimental data with CFD. • We also benchmarked the 4 × 4 HCF mixing experimental data with CFD. • We derived new friction factors for PWR and BWR designs at PWR and BWR operating conditions from CFD. • We showed the importance of modeling the 3D conduction in HCF in steady state and transient conditions. - Abstract: In order to increase the power density of current and new light water reactor designs, the helical cruciform fuel (HCF) rods have been proposed. The HCF rod is equivalent to a thin cylindrical rod, with 4 fuel containing vanes, wrapped around it. The HCF rods increase the surface area to volume ratio of the fuel and enhance the inter-subchannel mixing due to their helical shape. The rods do not need supporting grids, as they are packed to periodically contact their neighbors along the flow direction, enabling a higher power density in the core. The HCF rods were reported to have the potential to uprate existing PWRs by 45% and BWRs by 20%. In order to quantify the mixing behavior of the HCF rods based on their twist pitch, experiments were previously performed at atmospheric pressures with single phase water in a 4 by 4 HCF and cylindrical rod bundles. In this paper, the experimental results on pressure drop and mixing are benchmarked with computational fluid dynamic (CFD) using steady state the Reynolds average Navier–Stokes (RANS) turbulence model. The sensitivity of the CFD approach to computational domain, mesh size, mesh shape and RANS turbulence models are examined against the experimental conditions. Due to the refined radial velocity profile from the HCF rods twist, the turbulence models showed little sensitivity to the domain. Based on the CFD simulations, the total pressure drops under the PWR and BWR conditions are expected to be about 10% higher than the values previously reported solely from an empirical correlation based on the

  5. Upper end plug of fuel rod

    International Nuclear Information System (INIS)

    The present invention concerns a seal-welding of an upper end plug of a fuel rod for nuclear fuels conducted in a final stage of molding fabrication of the fuel rod in a pressurized helium gas. A welding protrusion is formed at the periphery of a vent hole on the upper surface of the upper end plug, and the welding protrusion is melted by irradiation of laser beams. The melted protrusion intrudes into the end portion of the bent hole by capillary to close the vent hole. The upper end plug can be closed by an extremely simple operation of irradiating the laser beams to the protrusion. Control for electrode gap on every fuel rods and exchange for the electrodes as in TIG welding can be saved, thereby enabling to speed up and simplify the sealing operation for the upper end plug. (N.H.)

  6. Spacer for fuel rods in nuclear fuel elements

    International Nuclear Information System (INIS)

    Spacers for fuel rods in nuclear reactor fuel elements are described, especially for use aboard ships. Spacers are used in a grid formed by web plates orthogonally intersecting and assembled together in a tooth-comb fashion forming a plurality of channels. The web plates are joined together and each of the web plates includes apertures through which resilient and separator members are joined. The resilient and separator members are joined. The resilient and separator members are in adjacent channels and with other similar members in the same channel, contact a fuel rod in the channel. The contact pressure between the members and fuel rod is radially directed

  7. Film cooling of vertical fuel rods

    International Nuclear Information System (INIS)

    Spray cooling of vertical rods has been studied at low heat fluxes appropriate to the removal of fission product heating following a reactor shut down. A series of tests have been made at atmospheric pressure using electrically heated rods, both singly and in a seven rod cluster, cooled by a falling film of water. Four modes of film breakdown were observed; progressive evaporation of the film; dry-patch formation due to surface tension effects at high inlet subcooling; stripping of the film by the flooding action of counterflow steam; and the disruption of the film on a hot rod caused by sputtering. Each of these phenomena is described in relation to the application of film cooling to long vertical fuel rod clusters. (author)

  8. Development of neural network for analysis of local power distributions in BWR fuel bundles

    International Nuclear Information System (INIS)

    A neural network model has been developed to learn the local power distributions in a BWR fuel bundle. A two layers neural network with total 128 elements is used for this model. The neural network learns 33 cases of local power peaking factors of fuel rods with given enrichment distribution as the teacher signals, which were calculated by a fuel bundle nuclear analysis code based on precise physical models. This neural network model studied well the teacher signals within 1 % error. It is also able to calculate the local power distributions within several % error for the different enrichment distributions from the teacher signals when the average enrichment is close to 2 %. This neural network is simple and the computing speed of this model is 300 times faster than that of the precise nuclear analysis code. This model was applied to survey the enrichment distribution to meet a target local power distribution in a fuel bundle, and the enrichment distribution with flat power shape are obtained within short computing time. (author)

  9. Analysis of Double-encapsulated Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Medvedev, Pavel G [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Perez, Danielle Marie [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

    2014-09-01

    In an LWR fuel rod, the cladding encapsulates the fuel, contains fission products, and transfers heat directly to the water coolant. In some situations, it may be advantageous to separate the cladding from the coolant through use of a secondary cladding or capsule. This may be done to increase confidence that the fuel or fission products will not mix with the coolant, to provide a mechanism for controlling the rod temperature, or to place multiple experimental rodlets within a single housing. With an axisymmetric assumption, it is possible to derive closed-form expressions for the temperature profile in a fuel rod using radially-constant thermal conductivity in the fuel. This is true for both a traditional fuel-cladding rod and a double-encapsulated fuel (fuel, cladding, capsule) configuration. Likewise, it is possible to employ a fuel performance code to analyse both a traditional and a double-encapsulated fuel. In the case of the latter, two sets of gap heat transfer conditions must be imposed. In this work, we review the equations associated with radial heat transfer in a cylindrical system, present analytic and computational results for a postulated power and gas mixture history for IFA-744, and describe the analysis of the AFC-2A, 2B metallic fuel alloy experiments at the Advanced Test Reactor, including the effect of a release of fission products into the cladding-capsule gap. The computational results for these two cases were obtained using BISON, a fuel performance code under development at Idaho National Laboratory.

  10. Development of neural network for predicting local power distributions in BWR fuel bundles considering burnable neutron absorber

    International Nuclear Information System (INIS)

    A neural network model is under development to predict the local power distribution in a BWR fuel bundle as a high speed simulator of precise nuclear physical analysis model. The relation between 235U enrichment of fuel rods and local peaking factor (LPF) has been learned using a two-layered neural network model ENET. The training signals used were 33 patterns having considered a line symmetry of a 8x8 assembly lattice including 4 water rods. The ENET model is used in the first stage and a new model GNET which learns the change of LPFs caused by burnable neutron absorber Gadolinia, is added to the ENET in the second stage. Using this two-staged model EGNET, total number of training signals can be decreased to 99. These training signals are for zero-burnup cases. The effect of Gadolinia on LPF has a large nonlinearity and the GNET should have three layers. This combined model of EGNET can predict the training signals within 0.02 of LPF error, and the LPF of a high power rod is predictable within 0.03 error for Gadolinia rod distributions different from the training signals when the number of Gadolinia rods is less than 10. The computing speed of EGNET is more than 100 times faster than that of a precise nuclear analysis model, and EGNET is suitable for scoping survey analysis. (author)

  11. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  12. Calibration of the TVO spent BWR reference fuel assembly

    International Nuclear Information System (INIS)

    In 1989 the Support Programmes of Finland (FSP) and Sweden (SSP) initiated a joint task to cross calibrate the burnup of the IAEA spent BWR reference fuel assembly at the TVO AFR storage facility (TVO KPA-STORE) in Finland. The reference assembly, kept separately under the IAEA seal, is used for verification measurements of spent fuel by GBUV method (SG-NDA-38). The cross calibration was performed by establishing a calibration curve, 244Cm neutron rate versus burnup, using passive neutron assay (PNA) measurements. The declared burnup of the reference assembly was compared with the burnup value deduced from the calibration curve. A calibration line was also established by using the GBUV method with the aid of high resolution gamma ray spectrometry (HRGS). Normalization between the two different facilities was performed using sealed neutron and gamma calibration sources. The results of the passive neutron assay show consistency, better than 1 %, between the declared mean burnup of the reference assembly and the burnup deduced from the calibration curve. The corresponding consistency is within +-2 % for the HRGS measurements

  13. Study of transient rod extraction failure without RBM in a BWR; Estudio del transitorio error de extraccion de barra sin RBM en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

  14. Qualification of helium measurement system for detection of fuel failures in a BWR

    Science.gov (United States)

    Larsson, I.; Sihver, L.; Loner, H.; Grundin, A.; Helmersson, J.-O.; Ledergerber, G.

    2014-05-01

    There are several methods for surveillance of fuel integrity during the operation of a boiling water reactor (BWR). The detection of fuel failures is usually performed by analysis of grab samples of off-gas and coolant activities, where a measured increased level of ionizing radiation serves as an indication of new failure or degradation of an already existing one. At some nuclear power plants the detection of fuel failures is performed by on-line nuclide specific measurements of the released fission gases in the off-gas system. However, it can be difficult to distinguish primary fuel failures from degradation of already existing failures. In this paper, a helium measuring system installed in connection to a nuclide specific measuring system to support detection of fuel failures and separate primary fuel failures from secondary ones is presented. Helium measurements provide valuable additional information to measurements of the gamma emitting fission gases for detection of primary fuel failures, since helium is used as a fill gas in the fuel rods during fabrication. The ability to detect fuel failures using helium measurements was studied by injection of helium into the feed water systems at the Forsmark nuclear power plant (NPP) in Sweden and at the nuclear power plant Leibstadt (KKL) in Switzerland. In addition, the influence of an off-gas delay line on the helium measurements was examined at KKL by injecting helium into the off-gas system. By using different injection rates, several types of fuel failures with different helium release rates were simulated. From these measurements, it was confirmed that the helium released by a failed fuel can be detected. It was also shown that the helium measurements for the detection of fuel failures should be performed at a sampling point located before any delay system. Hence, these studies showed that helium measurements can be useful to support detection of fuel failures. However, not all fuel failures which occurred at

  15. Radial optimization of a BWR fuel cell using genetic algorithms

    International Nuclear Information System (INIS)

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U235 and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix the placement of

  16. Behavior of BWR-type fuel elements with B{sub 4}C/steel absorber tested under severe fuel damage conditions in the CORA facility

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L.; Hagen, S.; Hofmann, P.; Schanz, G.

    2009-01-15

    The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program was to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations were tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The BWR-type bundles consisted of 18 fuel rod simulators (heated and unheated rods), an absorber blade of steel containing eleven absorber rods filled with boron carbide powder. The larger bundle CORA-18 contained the same number of absorber rods but was made up of 48 fuel rod simulators. All BWR bundles were surrounded by a zircaloy shroud and the absorber blades by a channel box wall on each side, also made of zircaloy. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence was simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. In all experiments bundle destruction started in the upper region (axially) with melting of the absorber blade and the absorber rod cladding at about 1250 C by interaction of boron carbide and steel. After destruction of the channel box walls this melt attacked the zircaloy fuel rod cladding and started to interact with the UO{sub 2} pellets. The test bundles also resulted in severe oxidation of the following components made of zircaloy: shroud, cladding, and grid spacers at the central and upper positions. Relocated absorber melt

  17. Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1998-01-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  18. Validation of SCALE (SAS2H) isotopic predictions for BWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; DeHart, M.D.

    1998-09-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  19. Implementation of a methodology to perform the uncertainty and sensitivity analysis of the control rod drop in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. del C.

    2015-07-01

    A methodology to perform uncertainty and sensitivity analysis for the cross sections used in a Trace/PARCS coupled model for a control rod drop transient of a BWR-5 reactor was implemented with the neutronics code PARCS. A model of the nuclear reactor detailing all assemblies located in the core was developed. However, the thermohydraulic model designed in Trace was a simple model, where one channel representing all the types of assemblies located in the core, it was located inside a simple vessel model and boundary conditions were established. The thermohydraulic model was coupled with the neutronics model, first for the steady state and then a Control Rod Drop (CRD) transient was performed, in order to carry out the uncertainty and sensitivity analysis. To perform the analysis of the cross sections used in the Trace/PARCS coupled model during the transient, Probability Density Functions (PDFs) were generated for the 22 parameters cross sections selected from the neutronics parameters that PARCS requires, thus obtaining 100 different cases for the Trace/PARCS coupled model, each with a database of different cross sections. All these cases were executed with the coupled model, therefore obtaining 100 different outputs for the CRD transient with special emphasis on 4 responses per output: 1) The reactivity, 2) the percentage of rated power, 3) the average fuel temperature and 4) the average coolant density. For each response during the transient an uncertainty analysis was performed in which the corresponding uncertainty bands were generated. With this analysis it is possible to observe the results ranges of the responses chose by varying the uncertainty parameters selected. This is very useful and important for maintaining the safety in the nuclear power plants, also to verify if the uncertainty band is within of safety margins. The sensitivity analysis complements the uncertainty analysis identifying the parameter or parameters with the most influence on the

  20. The BG18, a B(U)F type package used for the transport of irradiated fuel rods - return of experience

    Energy Technology Data Exchange (ETDEWEB)

    Juergen, S.; Herman, S. [Transnubel, Dessel (Belgium)

    2004-07-01

    The purpose of this presentation is to share the return of experience of Transnubel after a period of nearly 3 years operation of the BG18 package in several nuclear power plants and hot cell facilities. This package has been used mainly for the shipment of full scale as well as samples of irradiated fuel rods - UOX or MOX, PWR or BWR.

  1. Behavior of water reactor fuel rod

    International Nuclear Information System (INIS)

    This paper reviewed the fuels used widely in forms of (1) Zircaloy-sheathed UO2 fuel in light water-commercial power reactor, (2) Zircaloy-sheathed PuO2-UO2 fuel in plutonium-thermal reactor and advanced reactor (ATR), (3) aluminide and silicide fuel in Material Testing Reactors. From fundamental view points, physical/chemical properties and irradiation behaviors of both fuels and zircaloy claddings are briefly reviewed in chapters 1 and 2. Change of the fuel rod physical parameters with progress of burn-up are summed up in chapter 3. Some fuel troubles and failures encountered in past usage of worldwide LWR fuels are introduced with counterplans taken. In the last session of this chapter, recent results of R and D works have been carried out by fuel vendors are reviewed. Especially, in-core behaviors of PCI-remedy fuels developed to use for high burn-up extension and for load-follow operation are highlighted. Reactor accidents occurred through past forty years are surveyed and reviewed. Fuel behaviors during the reactivity initiated accident (RIA), the power-coolant mismatch (PCM), and the loss-of-coolant accident (LOCA) are taken into this review by using disclosed literatures. Safety criteria being used in Japanese licensing authorities are introduced relating to the fuel design limit. (author)

  2. Fuel rod design by statistical methods for MOX fuel

    International Nuclear Information System (INIS)

    Statistical methods in fuel rod design have received more and more attention during the last years. One of different possible ways to use statistical methods in fuel rod design can be described as follows: Monte Carlo calculations are performed using the fuel rod code CARO. For each run with CARO, the set of input data is modified: parameters describing the design of the fuel rod (geometrical data, density etc.) and modeling parameters are randomly selected according to their individual distributions. Power histories are varied systematically in a way that each power history of the relevant core management calculation is represented in the Monte Carlo calculations with equal frequency. The frequency distributions of the results as rod internal pressure and cladding strain which are generated by the Monte Carlo calculation are evaluated and compared with the design criteria. Up to now, this methodology has been applied to licensing calculations for PWRs and BWRs, UO2 and MOX fuel, in 3 countries. Especially for the insertion of MOX fuel resulting in power histories with relatively high linear heat generation rates at higher burnup, the statistical methodology is an appropriate approach to demonstrate the compliance of licensing requirements. (author)

  3. Fuel rod behaviour during transients

    International Nuclear Information System (INIS)

    The clad deformation code CANSWELL-2 is described. This is used, either as a stand-alone code or within MABEL-2, to predict and analyse the results of LOCA simulations in the Halden and NRU reactors and in the KfK and PROPAT rigs. Experimental evidence on fuel behaviour in RIA, PCM and ATWS events is presented with inclusion of certain FRAP-T5 results. Published calculations from the accident codes FRAP-T4 and FRAP-T5 are compared with experimental results in simulated loss of coolant tests in the Power Burst Facility. The limitations of this code in its treatment of RIA, PCM and ATWS events are considered. (U.K.)

  4. Experimental design for HTGR fuel rods

    International Nuclear Information System (INIS)

    Fuel rods for the high temperature gas cooled reactor are composed of pyrolytic carbon coated fuel particles bounded by a carbonaceous matrix. Because of differential shrinkage between coated particles and the carbonaceous matrix, breakage of the pyrolytic coating has been observed with certain combinations of coated particles and matrix compositions. The pyrolytic coating is intended to be the primary containment for fission products. Therefore, an experiment is desired to determine the breakage characteristics of different strength coated particles combined with different matrix compositions during irradiation

  5. Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Adamsson, Carl, E-mail: carl.adamsson@psi.ch [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden); Le Corre, Jean-Marie, E-mail: lecorrjm@westinghouse.com [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden)

    2011-08-15

    Highlights: > The MEFISTO code efficiently and accurately predicts the dryout event in a BWR fuel bundle, using a mechanistic model. > A hybrid approach between a fast and robust sub-channel analysis and a three-field two-phase analysis is adopted. > MEFISTO modeling approach, calibration, CPU usage, sensitivity, trend analysis and performance evaluation are presented. > The calibration parameters and process were carefully selected to preserve the mechanistic nature of the code. > The code dryout prediction performance is near the level of fuel-specific empirical dryout correlations. - Abstract: Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle

  6. Thermal phenomenae in nuclear fuel rods

    International Nuclear Information System (INIS)

    Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor)

  7. High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

  8. Analysis by the Monte Carlo method of doses around the pool of storage of the control rods irradiated in a BWR reactor; Analisis mediante el metodo de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactror BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas, J.; Gallardo, S.

    2011-07-01

    The control rods of a boiling water reactor (BWR) are subject to a neutron flux and thus become activated during their stay in the reactor core. Activation occurs especially in the stainless steel components and impurities. The activity generated results in a dose around the bar, while it le unimportant in the reactor, but to be taken into account when removed f ron it. The bars drawn are stored on hangers placed in the storage pools of spent fuel f ron the plant. Each hanger 12 accommodates control rods and are arranged so that at least three meters of water abode the heads of the control rods. The dose received by potentially exposed workers who are in the vicinity of the storage must be calculated to ensure adequate protection of the came. This dose can be decreased significantly by changing the arrangement of the bars on hangers.

  9. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO2 fuel in Japan. The design concept should be compatible with UO2 fuel design. High burnup UO2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO2 8 x 8 array fuel developed for a second step of UO2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  10. Revaluation on measured burnup values of fuel assemblies by post-irradiation experiments at BWR plants

    International Nuclear Information System (INIS)

    Fuel composition data for 8x8 UO2, Tsuruga MOX and 9x9-A type UO2 fuel assemblies irradiated in BWR plants were measured. Burnup values for measured fuels based on Nd-148 method were revaluated. In this report, Nd-148 fission yield and energy per fission obtained by burnup analyses for measured fuels were applied and fuel composition data for the measured fuel assemblies were revised. Furthermore, the adequacies of revaluated burnup values were verified through the comparison with burnup values calculated by the burnup analyses for the measured fuel assemblies. (author)

  11. Process for tightly sealing nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    This invention refers to a process for pressurising and tightly sealing fuel rods used in nuclear reactors. The fuel rods utilised in commercial nuclear reactors are usually composed of a zircaloy tubular cladding of around 15 mm in diameter and up to 5 m long, filled with fuel pellets maintained in place by end plugs sealed on each end of the cladding. The main purpose of the invention is to promote a process using laser beam welding equipment to seal the end plugs on the fuel rods, drill a pressurising hole in one of the end plugs and hermetically seal the hole after the fuel rod has been placed under pressure

  12. Nondestructive assay of green HTGR fuel rods

    International Nuclear Information System (INIS)

    This report describes the nondestructive (NDA) work done at Los Alamos during 1979 and 1980 as part of the New Brunswick Laboratory-sponsored evaluation of NDA of the uranium content of fabricated fuel rods for high-temperature gas-cooled reactors (HTGR). The methods used (delayed neutron and passive gamma ray) are concisely described, and the results are summarized and compared in graphical and tabular form. The results indicate that, with the use of proper physical standards, accuracies within about 1 percent should be achievable by NDA procedures

  13. Concept of the core for a small-to-medium-sized BWR that does not use control rods during normal operation

    International Nuclear Information System (INIS)

    A small-to-medium-sized boiling water reactor (BWR) with a natural circulation system is being developed for countries where initial investment funds for construction are limited and electricity transmission networks have not been fully constructed. To lighten operators' work load, a core that does not use control rods during normal operation (control rod-free core) was developed by using a neutronics calculation system coupled with core flow evaluation. The control rod-free core had large core power fluctuation with conventional burnable poison design. The target of core power fluctuation was set to less than 10% and was achieved by optimization of burnable poison arrangement. (author)

  14. Conceptual design report of the SMART fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    The SMART fuel rod is based on 17 x 17 KOFA(Korea Fuel Assembly) fuel rod of the 950MWe pressurize water reactor. The fuel stack length of the KOFA is 3658mm, otherwise SMART fuel rod stack length is 2000mm. The fuel rod contains UO{sub 2} pellets with the enrichment of 4.95%. All the fuel in core will be replaced every 35 months. The average LHGR of the fuel rod is 120 W/cm, commercial PWR is 178 W/cm, SMART LHGR is lower about 31% than commercial PWR. The core inlet and outlet temperature of coolant are respectively 270 deg C and 310 deg C, commercial PWR are respectively 291.6 deg C and 326.8 deg C, SMART inlet and outlet temperature is lower averaged 19.2 deg C than commercial PWR. The coolant use mixed soluble ammonia in high purity water and boron is not in. The general performance of the fuel rod UO{sub 2} pellet has been already verified through the sufficient burnup (60,000 MWd/MTU-rod avg.) experience as the rods of same design in commercial PWR's. But cladding corrosion is required the further verification. (author). 13 refs., 3 figs., 8 tabs.

  15. Radial distribution of UO2 and Gd2O3 in fuel cells of a BWR Reactor

    International Nuclear Information System (INIS)

    The fuel system that is used at the moment in a power plant based on power reactors BWR, includes as much like the one of its substantial parts to the distribution of the fissile materials like a distribution of burnt poisons within each one of the cells which they constitute the fuel assemblies, used for the energy generation. Reason why at the beginning of a new operation cycle in a reactor of this type, the reactivity of the nucleus should be compensated by the exhaustion of the assemblies that it moves away of the nucleus for their final disposition. This compensation is given by means of the introduction of the recharge fuel, starting from the UO2 enriched in U235, and of the Gadolinium (Gd2O3). The distribution of these materials not only defines the requirements of energy generation, but in certain measures also the form in that the margins will behave to the limit them thermal during the operation of the reactor. These margins must be taken into account for the safe and efficient extraction of the energy of the fuel. In this work typical fuel cells appear that are obtained by means of the use of a emulation model of an ants colony. This model allows generating from a possible inventory of values of enrichment of U235, as well as of concentration of Gadolinium a typical fuel cell, which consists of an arrangement of lOxlO rods, of which 92 contain U235, some of these rods contain a concentration of Gd2O3 and 8 of the total contain only water. The search of each cell finishes when the value of the Local Peak Power Factor (LPPF) in the cell reaches a minimal value, or when a pre established value of iterations is reached. The cell parameters are obtained from the results of the execution of the code HELIOS, which incorporates like a part integral of the search algorithm. (Author)

  16. Thermal behavior simulation of a nuclear fuel rod through an eletrically heated rod

    International Nuclear Information System (INIS)

    In thermalhydraulic loops the nuclear industry often uses electrically heated rods to simulate power transients, which occur in nuclear fuel rods. The development and design of a electrically heated rod, by supplying the dimensions and materials which should be used in order to yeld the same temperature and heat flux at the surfaces of the nuclear rod and the electrically heated rod are presented. To a given nuclear transient this equality was obtained by fitting the linear power through the lumped parameters technique. (Author)

  17. Common Cause Failure Analysis of Control Rods and Drives in the Swedish and Finnish BWR Plants. Operating Experiences in 1983 - 2003

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, Tuomas [Avaplan Oy, Espoo (Finland)

    2006-11-15

    The control rod and drives in a Boiling Water Reactor (BWR) constitute a highly redundant system. The reliability of the system is determined by how well the design withstands dependencies, as Common Cause Failures (CCFs). This report upgrades an earlier data collection on CCFs of control rod and drives (SKI Report 1996:77) to more recent years, with the objective to report the data to ICDE project (International Common Cause Failure Data Exchange) and to the safety analysts in the Nordic countries. The operating experiences were analyzed at the BWRs of former Asea-Atom design, comprising 9 units in Sweden and Olkiluoto 1 and 2 in Finland, covering years 1983 - 2003. A new logical scheme was developed to classify interconnected failure modes of the two redundant functions for reactivity shutdown, fast hydraulic insertion and slower screw insertion of control rods. The scheme makes an explicit distinction between the different attributes of the failure event: - affected function - affected movement direction - detectability - criticality, i.e. inoperable control rod function versus only degraded functionality Another novel idea emerged for grouping the events according to generic failure mechanism. The generic classes will help to organize and structure the information efficiently, because in most cases within a class, the failure modes prove to be same, or there are only a few alternatives to chose from. From the set of 72 candidate cases, altogether 27 actual or more significant potential CCFs were screened out. Special emphasis was placed to identify any multiple failure or degradation indicating that adjacent rods would be more vulnerable to failure, because such phenomena are far more critical for reactivity shutdown as compared to failure of randomly placed rods. Only slight tendency of position dependence could be determined. Another positive insight is that the events, where foreign objects caused the jamming of rod insertion, were separated by both

  18. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

  19. Welding nuclear reactor fuel rod end plugs

    International Nuclear Information System (INIS)

    Apparatus for applying a vacuum to a nuclear fuel rod cladding tube's interior through its open end while girth welding an inserted end plug to its other end. An airtight housing has an orifice with a seal which can hermetically engage the tube's open end. A vacuum hose has one end connected to the housing and the other end connected to a vacuum pump. A mechanized device which moves the housing to engage or disengage its seal with the tube's open end includes at least one arm having one end attached to the housing and the other end pivotally attached to a movable table; an arm rotating device to coaxially align the housing's orifice with the welding-positioned tube; and a table moving device to engage the seal of the coaxially aligned orifice with the tube's open end. (author)

  20. Welding nuclear reactor fuel rod end plugs

    International Nuclear Information System (INIS)

    Apparatus for applying a vacuum to a nuclear fuel rod cladding tube's interior through its open end while girth welding an inserted end plug to its other end. An airtight housing has an orifice with a seal which can hermetically engage the tube's open end. A vacuum hose has one end connected to the housing and the other end connected to a vacuum pump. A mechanized device moves the housing to engage or disengage its seal with the tube's open end. Preferably the mechanized device includes an arm having one end attached to the housing and the other end pivotally attached to a moveable table; an arm rotating device to coaxially align the housing's orifice with the welding-positioned tube; and a table moving device to engage the seal of the coaxially aligned orifice with the tube's open end

  1. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  2. Fuel performance annual report for 1981. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.; Tokar, M.

    1982-12-01

    This annual report, the fourth in a series, provides a brief description of fuel performance during 1981 in commercial nuclear power plants. Brief summaries of fuel operating experience, fuel problems, fuel design changes and fuel surveillance programs, and high-burnup fuel experience are provided. References to additional, more detailed information and related NRC evaluations are included.

  3. Development of a fuel rod thermal-mechanical analysis code for high burnup fuel

    International Nuclear Information System (INIS)

    The thermal-mechanical analysis code for high burnup BWR fuel rod has been developed by NFI. The irradiation data accumulated up to the assembly burnup of 55 GWd/t in commercial BWRs were adopted for the modeling. In the code, pellet thermal conductivity degradation with burnup progress was considered. Effects of the soluble FPs, irradiation defects and porosity increase due to RIM effect were taken into the model. In addition to the pellet thermal conductivity degradation, the pellet swelling due to the RIM porosity was studied. The modeling for the high burnup effects was also carried out for (U, Gd)O2 and MOX fuel. The thermal conductivities of all pellet types, UO2, (U, Gd)O2 and (U, Pu)O2 pellets, are expressed by the same form of equation with individual coefficient γ in the code. The pellet center temperature was calculated using this modeling code, and compared with measured values for the code verification. The pellet center temperature calculated using the thermal conductivity degradation model agreed well with the measured values within ±150 deg. C. The influence of rim porosity on pellet center temperature is small, and the temperature increase in only 30 deg. C at 75 GWd/t and 200 W/cm. The pellet center temperature of MOX fuel was also calculated, and it was found that the pellet center temperature of MOX fuel with 10wt% PuO2 is about 60 deg. C higher than UO2 fuel at 75 GWd/t and 200 W/cm. (author)

  4. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  5. A methodology for obtaining the control rod patterns in a BWR using genetic algorithms

    International Nuclear Information System (INIS)

    In this work the GACRP system based on the genetic algorithms technique for the obtaining of the drivers of control bars in a BWR reactor is presented. This methodology was applied to a transition cycle and a one of balance of the Laguna Verde nuclear power station (CNLV). For each one of the studied cycles, it was executed the methodology with a fixed length of the cycle and it was compared the effective multiplication factor of neutrons at the end of the cycle that it is obtained with the proposed drivers of control bars and the multiplication factor of neutrons obtained by means of a Haling calculation. It was found that it is possible to extend several days the length of both cycles with regard to the one Haling calculation. (Author)

  6. MELCOR/SNAP analysis of Chinshan (BWR/4) Nuclear Power Plant spent fuel pool for the similar Fukushima accident

    International Nuclear Information System (INIS)

    Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP event occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool, by using MELCOR 2.1 and SNAP 2.2.7 codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP spent fuel pool (SFP). There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP MELCOR/SNAP model. And the transient analysis under the SFP cooling system failure condition was performed. Besides, in order to study the detailed thermal-hydraulic performance of this transient, TRACE was used in this analysis. CFD data from INER report was used to compare with the results of MELCOR and TRACE. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Besides, the animation model of Chinshan NPP SFP was presented using the animation function of SNAP with MELCOR analysis results. (author)

  7. SEFLEX fuel rod simulator effects in flooding experiments. Pt. 1

    International Nuclear Information System (INIS)

    The aim of the SEFLEX program has been to quantify the influence of the design and the physical properties of different fuel rod simulators on heat transfer and quench front progression in unblocked and blocked rod bundles during the reflood phase of a LOCA in a PWR. Fuel rod simulators with Zy claddings and a gas-filled gap between claddings and pellets exhibit lower peak cladding temperatures and shorter quench times than gapless heater rods with stainless steel claddings. Grid spacers cause significant cooling enhancement downstream during the time span at which maximum cladding temperatures occur. Ballooned Zy claddings, forming e.g. a 90 percent blockage, are quenched substantially earlier than thickwall stainless steel blockage sleeves attached to the rods, and even earlier than undeformed rod claddings. A comparison of test data with results of the 'Best Estimate' computer program COBRA-TF shows a good agreement with unblocked bundle data including grid spacer effects. (orig./HP)

  8. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor; Evaluacion del diseno radial de celdas de combustible en un ciclo de operacion de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez C, J.; Martin del Campo M, C. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jgco@ver.megared.net.mx

    2003-07-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  9. Evaluation of fission gas release in Oskarshamn 2 fuel rods

    International Nuclear Information System (INIS)

    Data on fission gas release from 30 Oskarshamn 2 fuel rods is presented and analyzed. These rods were selected by gamma scanning screening measurements performed on 65 fuel rods. The rods experienced peak power levels 25-34 kW/m in the burnup range of 18.5 to 41 MWd/kg U, higher power and burnups than the previous experimental programs, the Oskarshamn 1 and Barsebeck 1 rods. As a result, a larger fraction of rods fall into the high fission gas release group. The STAV5 code is used for the analysis of fission gas release data. The code is capable in most instances to capture the experimental points. It takes as an input the fuel rod design dimensions data and power histories calculated by other ASEA-ATOM codes. With STAV5, one is enabled to treat each rod individually on the bases of its detailed power histories and its UO2 stability properties. The report includes manufacturing data, PIE data, and the description of the models used in STAV5 for gas release prediction. The results of STAV5 calculations for a number of rods are reported and several remarks for improvements have been proposed. (author)

  10. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    International Nuclear Information System (INIS)

    utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

  11. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

  12. Experience of Areva in fuel services for PWR and BWR

    International Nuclear Information System (INIS)

    AREVA being an integrated supplier of fuel assemblies has included in its strategy to develop services and solutions to customers who desire to improve the performance and safety of their fuel. These services go beyond the simple 'after sale' services that can be expected from a fuel supplier: The portfolio of AREVA includes a wide variety of services, from scientific calculations to fuel handling services in a nuclear power plant. AREVA is committed to collaborate and to propose best-in-class solutions that really make the difference for the customer, based on 40 years of Fuel design and manufacturing experience. (Author)

  13. Comparison of metaheuristic optimization techniques for BWR fuel reloads pattern design

    International Nuclear Information System (INIS)

    Highlights: ► This paper shows a performance comparison of several optimization techniques for fuel reload in BWR. ► Genetic Algorithms, Neural Networks, Tabu Search and several Ant Algorithms were used. ► All optimization techniques were executed under same conditions: objective function and an equilibrium cycle. ► Fuel bundles with minor actinides were loaded into the core. ► Tabu search and Ant System were the best optimization technique for the studied problem. -- Abstract: Fuel reload pattern optimization is a crucial fuel management activity in nuclear power reactors. Along the years, a lot of work has been done in this area. In particular, several metaheuristic optimization techniques have been applied with good results for boiling water reactors (BWRs). In this paper, a comparison of different metaheuristics: genetic algorithms, tabu search, recurrent neural networks and several ant colony optimization techniques, were applied, in order to evaluate their performance. The optimization of an equilibrium core of a BWR, loaded with mixed oxide fuel composed of plutonium and minor actinides, was selected to be optimized. Results show that the best average values are obtained with the recurrent neural networks technique, meanwhile the best fuel reload was obtained with tabu search. However, according to the number of objective functions evaluated, the two fastest optimization techniques are tabu search and Ant System.

  14. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.)

  15. Investigation of Burnup Credit Issues in BWR Fuel

    International Nuclear Information System (INIS)

    Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents, followed by criticality calculations to assess the value of keff for a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel

  16. Experience of MOX-fuel operation in the Gundremmingen BWR plant: Nuclear characteristics and in-core fuel management

    International Nuclear Information System (INIS)

    After 4 years of good experience with MOX-fuel operation in the BWR plants Gundremmingen units B and C the number of inserted MOX-FAs will be increased in the future continuously. Until now all MOX-FAs are in good condition. Furthermore calculations and measurements concerning zero power tests and tip measurements are in good agreement as expected: all results lead to the conclusion that MOX-FAs can be calculated with the same precision as uranium-FAs. (author)

  17. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  18. Taking burnup credit for interim storage and transportation system for BWR fuels

    International Nuclear Information System (INIS)

    In order to establish a realistic burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of several modules such as TGBLA, ORIGEN, CITATION, MCNP, and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. A compact measurement system for a fuel assembly has been being developed to meet requirements for the burnup determination, the neutron emission-rate evaluation, and the nuclear materials management. For a spent MOX fuel, a neutron emission rate measurement method has been being developed. The system consists of Cd-Te detectors and / or fission chambers. Some model calculations were carried out for the latest design BWR fuels. The effect of taking burnup credit for a transportation cask is shown. (authors)

  19. Fuel performance improvement program. Quarterly/annual progress report, October 1977--September 1978. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Crouthamel, C.E. (comp.)

    1978-10-01

    This quarterly/annual report reviews and summarizes the activities performed in support of the Fuel Performance Improvement Program (FPIP) during Fiscal Year 1978 with emphasis on those activities that transpired during the quarter ending September 30, 1978. Significant progress has been made in achieving the primary objectives of the program, i.e., to demonstrate commercially viable fuel concepts with improved fuel - cladding interaction (FCI) behavior. This includes out-of-reactor experiments to support the fuel concepts being evaluated, initiation of instrumented test rod experiments in the Halden Boiling Water Reactor (HBWR), and fabrication of the first series of demonstration rods for irradiation in the Big Rock Point Reactor (BRPR).

  20. Failure development in leaking LWR fuel rods - a literature survey

    International Nuclear Information System (INIS)

    This review of failure development is based largely on a literature survey up by some personal interpretation and judgement where conflicting evidence appears to exist. A brief review of known causes of defects in LWR fuel rods is included. Primary defects in fuel rods may not be single, isolated, leak sites, PCI defects being a classical example where very many penetrating cracks can form in a single fuel rod. The only secondary defect mechanism is internal hydriding, although knock-on events can follow on the same, or neighbouring rods, from defects that cause and interference with local coolant flow. The major part of this survey comprises information on leak rates from typical defects, causes and rates of development of primary leak sites, meachanisms and rates of development of secondary hydride defects and analysis of activity release measurements in an attempt to characterise the number and severity of defects present

  1. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  2. Test plan for thermogravimetric analyses of BWR spent fuel oxidation

    International Nuclear Information System (INIS)

    Preliminary studies indicated the need for additional low-temperature spent fuel oxidation data to determine the behavior of spent fuel as a waste form for a tuffy repository. Short-term thermogravimetric analysis tests were recommended in a comprehensive technical approach as the method for providing scoping data that could be used to (1) evaluate the effects of variables such as moisture and burnup on the oxidation rate, (2) determine operative mechanisms, and (3) guide long-term, low-temperature oxidation testing. The initial test series studied the temperature and moisture effects on pressurized water reactor fuel as a function of particle and grain size. This document presents the test matrix for studying the oxidation behavior of boiling water reactor fuel in the temperature range of 140 to 225/degree/C. 17 refs., 7 figs., 3 tabs

  3. Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning; Perfiles axiales de quemado y fraccion de huecos para calculos de criticidad con credito al quemado para combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-07-01

    This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

  4. BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR

    International Nuclear Information System (INIS)

    The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

  5. Design and optimization of a fuel reload of BWR with plutonium and minor actinides; Diseno y optimizacion de una recarga de combustible de BWR con plutonio y actinidos menores

    Energy Technology Data Exchange (ETDEWEB)

    Guzman A, J. R.; Francois L, J. L.; Martin del Campo M, C.; Palomera P, M. A. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: maestro_juan_rafael@hotmail.com

    2008-07-01

    In this work is designed and optimized a pattern of fuel reload of a boiling water reactor (BWR), whose fuel is compound of uranium coming from the enrichment lines, plutonium and minor actinides (neptunium, americium, curium); obtained of the spent fuel recycling of reactors type BWR. This work is divided in two stages: in the first stage a reload pattern designs with and equilibrium cycle is reached, where the reload lot is invariant cycle to cycle. This reload pattern is gotten adjusting the plutonium content of the assembly for to reach the length of the wished cycle. Furthermore, it is necessary to increase the concentration of boron-10 in the control rods and to introduce gadolinium in some fuel rods of the assembly, in order to satisfy the margin approach of out. Some reactor parameters are presented: the axial profile of power average of the reactor core, and the axial and radial distribution of the fraction of holes, for the one reload pattern in balance. For the design of reload pattern codes HELIOS and CM-PRESTO are used. In the second stage an optimization technique based on genetic algorithms is used, along with certain obtained heuristic rules of the engineer experience, with the intention of optimizing the reload pattern obtained in the first stage. The objective function looks for to maximize the length of the reactor cycle, at the same time as that they are satisfied their limits related to the power and the reactor reactivity. Certain heuristic rules are applied in order to satisfy the recommendations of the fuel management: the strategy of the control cells core, the strategy of reload pattern of low leakage, and the symmetry of a quarter of nucleus. For the evaluation of the parameters that take part in the objective function it simulates the reactor using code CM-PRESTO. Using the technique of optimization of the genetic algorithms an energy of the cycle of 10834.5 MW d/tHM is obtained, which represents 5.5% of extra energy with respect to the

  6. Estimating PWR fuel rod failures throughout a cycle

    International Nuclear Information System (INIS)

    A fuel performance engineer requires good prediction models for fuel conditions to help assure that any fuel repair operation he may recommend for the next refueling outage will have a minimal impact on nuclear plant operation. For nearly two decades, simple equilibrium equations have been used to provide estimates of the number of failed fuel rods in a pressurized water reactor (PWR) core. The unknown parameter is the isotopic escape rate (upsilon), which is often assumed to be --1 X 10/sup -8//s for the release of /sup 131/I from a 3- to 4-m-long PWR rod. The use of this escape rate value will generally produce end-of-cycle (EOC) predictions that are accurate within a factor of --3. When applied at the time when fuel rods initially fail, such as early in a reactor cycle, however, the prediction obtained may overestimate the number of failed rods present by a factor of 10 or more. While a goal of Combustion Engineering's (C-E's) efforts on failed fuel prediction (FFP) models over the past decade has been to increase the accuracy of the EOC estimate, recent efforts have emphasized improving prediction capability for failed rods present early in a reactor cycle. The C-E approach to modeling iodine release from failed fuel rods is based on dynamic escape rate theory that is incorporated in the C-E IODYNE (for iodine dynamic evaluation) code. This theory has been empirically modified to account for specific observed time dependencies of the release rates for /sup 131/I and /sup 133/I from a failed rod. In a current version of IODYNE, four such factors have been included in the FFP model, as described in this paper

  7. Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

  8. FRANCO, Finite Element Method (FEM) Fuel Rod Analysis for Solid and Annular Configurations

    International Nuclear Information System (INIS)

    1 - Description of program or function: The FRANCO code is a quasi- static two-dimensional fuel rod analysis code, that calculates the fuel temperature and material deformation as a function of heat generation rate. Both solid and annular fuel configurations are modeled. 2 - Method of solution: FRANCO uses two-dimensional finite element theory and applications for mechanical deformation and heat conduction, and determines the temperature distribution from the fuel center to the coolant adjacent to the clad at a position along the fuel rod axis. FRANCO calculates the average temperature of each radial division, the nodal displacement, and strain and stress within the fuel pellet and clad. The principal stresses, which represent maximum and minimum stresses within an element, result from Mohr's circle relationship between normal stresses. FRANCO is capable of predicting the thermo-mechanical behavior in the radial direction of a single fuel rod for both boiling water reactors (BWR's) and pressurized water reactors (PWR's). The cross sectional plane geometry of fuel rod is modeled using three-node constant strain triangular finite elements, and both thermal and mechanical solutions are computed with the same finite element configurations. The local linear heat generation rate is modeled as a uniform heat source in a fuel pellet, and the coolant temperature and heat transfer coefficient are applied as known boundary conditions at the boundary of the cladding surface. The total load to form the global force vector consists of the thermal load that results from thermal expansion of the material and the mechanical load exerted by pressure. FRANCO assumes the fuel-cladding gap region to be conductive material in order to simplify the analysis, and this gap is simulated by either an open gap or a closed gap model. A time- dependent problem can be simulated by FRANCO using quasi-static analysis when time-dependent parameters are provided. FRANCO can treat a steady-state or

  9. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

  10. Liquid film thickness on fuel rod under high pressure and high temperature steam-water two phase flow

    International Nuclear Information System (INIS)

    This paper deals with behavior of liquid film on the fuel rod which is very important for the critical power prediction. In this study, the liquid film measurement device using an ultrasonic transducer has been developed and the liquid film thickness data has been obtained for a simulated BWR 4 x 4 rod bundle under 1 MPa condition. The cooling fluid is steam-water mixture and flow direction is vertical. Also the following results were obtained. Firstly, the liquid film thickness becomes thinner with increasing quality and the liquid film thickness is about 0.2 mm at 9.3% of quality. Secondary, the time change of liquid film thickness becomes smaller with increasing the quality. It was found that the change of liquid film thickness becomes more smoothly near the dryout condition. (author)

  11. Heating apparatus for single-fuel-rod experiments

    International Nuclear Information System (INIS)

    A single-fuel-rod heating apparatus was constructed for installation to Semi-Homogeneous Experimental Assembly (SHE); which is used to measure the reactivity temperature coefficient of a single fuel rod in verification of the precision of nuclear design of the VHTR (very high temperature reactor). The apparatus raises the temperature of a single fuel rod up to 7000C. A fuel rod is enclosed in a silica tube coiled with nichrome heater. The silica tube is confined in a zircalloy tube of which outer surfaces are cooled with air to remove the radiation heat. The zircalloy tube is then confined in a double-walled evacuated aluminum tube to prevent heat transfer to SHE core. The heating apparatus consists of evacuation, cooling, instrumentation, control and safety system, beside the heating tube. The heating tube is inserted in a space made by withdrawing a graphite matrix tube along the central axis of SHE core. The base of the heating tube is connected to an evacuation system, which is set on the table of the movable half of SHE. Performance of the heating apparatus shown by test operation with a single graphite rod are: 1. Single fuel rod: 24 mm in diameter, 2400 mm long. 2. Heating ability: The heating up of a single fuel rod to 7000C in 40 min with electric power 3 kW. 3. Cooling capacity: Blower in flow 1.4 m3/min at pressure 0.4 kg/cm2. 4. Heat leakage: lower than 20 W. (author)

  12. Studies on the mechanics of fuel rods and fuel subassemblies - studies on the simulation irradiation project

    International Nuclear Information System (INIS)

    This report includes eight single papers which are a report of activities performed at the IRT on the mechanics of fuel rods and fuel subassemblies and on the Simulation Irradiation Project. (orig./HP)

  13. Evaluation of the reduction of boron-10 in the control rods in the BWR of the Laguna Verde Central, through steady state calculations; Evaluacion de la reduccion del Boro-10 en las barras de control en los BWR de la CLV, mediante calculos en estado estacionario

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J.L.; Perusquia, R.; Hernandez, J.L.; Ramirez S, J.R. [Departamento de Sistemas Nucleares, ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    One of the more important aspects related with the safety and economy in the operation of a nuclear power reactor, it is without a doubt the control of the reactivity. During the normal operation of a reactor of boiling water (BWR-Boiling Water Reactor), the control of the reactivity in the nucleus it is strongly determined by the efficiency of the control rods. In the case of the Laguna Verde Nuclear power station (CNLV) the nucleus of the reactors has 109 control rods grouped in 4 sets. The CNLV at the moment uses the CCC method (Control Cell Core) in the design of the cycle. With this method only the A2 group is used for the control of the reactivity at full power. With the purpose of quantifying the effect of the decrease of the burnable poison (B{sub 4}C) of the control rods and in particular to the effect due to the postulated lost of 10% of Boron 10, it was carried out a series of calculations of the nucleus in stationary state by means of the system of HELIOS/CM-PRESTO codes. In this work the main derived results of these 3D simulations(three dimensions) of the reactors of the CNLV are presented. It was analyzed the one behavior of the infinite neutron multiplication factor (K{sub infinite}), at fuel assemble cell level used in an equilibrium cycle for the CNLV. It was also analyzed the effect in the shutdown margin (ShutDown Margin- SDM) in cold condition CZP (Cold Zero Power). Its are also included those results of the ARI cases (All Rods In) and SRO (Strong Rod Out). From the cases in condition HFP (Hot Full Power) the behavior of the effective multiplication factor (K{sub eff}) is presented. (Author)

  14. Application of ultrasonic echo technique to the measurement of 2-dimensional local instantaneous liquid film thickness on a simulated nuclear fuel rod

    International Nuclear Information System (INIS)

    A subchannel analysis for nuclear reactor core thermal hydraulics solves the basic conservation equations for mass, momentum and energy for each sub-channel, taking into account the effect of cross flows between neighboring subchannels. The three fluid model formulation is often considered for the analysis of droplet dispersed annular flow regions where the liquid film flow on the fuel rods is determined by a balance among entrainment and deposition rates of liquid droplets in the vapor flow and evaporation rate on a heated wall. The critical heat flux in a BWR is therefore characterized by film dryout. It is well known that in a typical BWR fuel rod assembly there are mechanical spacers placed at, for example, every 50 centimeters in axial direction to keep the rod clearance adequately which in turn result in flow obstacles consequently, and thus local time dependent film flow structure is changed before and after these spacers. The previous studies revealed a general trend that the liquid film thickness normally reduces in a spacer region due to droplet entrainment rate enhanced by an accelerated vapor flow, whereas in a region downstream of the spacers it recovers again because of highly agitated nature of the flow behind the spacers and increased deposition rate. Thus the accurate prediction of time dependent 2-dimensional film thickness on the fuel rods is extremely indispensable for the evaluation of the critical heat flux for BWR fuel assemblies. However, for the moment, we do not have reliable data-base for non-homogeneous liquid film flow on the fuel rods mainly because of a lack in accurate non-intrusive measuring techniques. The purpose of this work is therefore to present the outline of a new development in high speed ultrasonic echo technique which we are now pursuing at our laboratory with a view to applying it to measurement of time-sequential 2-dimensional thickness of the film flow around a simulated nuclear fuel rod. The scanning time required to

  15. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  16. Optimization of fuel reloads for a BWR using the ant colony system

    International Nuclear Information System (INIS)

    In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

  17. On the FCMI concern of SFR fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Kyu; Lee, Hyun Seung; Yoon, Kyung Ho; Cheon, Jin Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The analysis procedure follows what has been presented but a finite element analysis result for the SFR fuel rods are included in this paper and thus an actual stress and the von Mises stress values are to be shown. It is that an FCMI failure would not be a big concern for the present SFR fuels because the stress components are bounded when a friction coefficient of the slug and cladding interface exceeds 0.3.

  18. CONDOR: neutronic code for fuel elements calculation with rods

    International Nuclear Information System (INIS)

    CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)

  19. TMI-2 standing fuel rod segments: Preliminary examination report

    Energy Technology Data Exchange (ETDEWEB)

    Akers, D W; Russell, M L

    1987-08-01

    The principal purpose of the nondestructive examination of the six fuel rod segments obtained from partially intact fuel assemblies still standing in the core of the Three Mile Island Unit 2 (TMI-2) reactor was to evaluate their potential for providing information that may contribute significantly toward understanding the TMI-2 accident. Objectives of the examinations were to (1) evaluate the necessity for performing destructive metallurgical and radiochemical examinations and (2) identify the best locations for removing samples for other analyses.

  20. Estimation and control in HTGR fuel rod fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Downing, D J; Bailey, M J

    1980-01-01

    A control algorithm has been derived for a HTGR Fuel Rod Fabrication Process utilizing the method of Box and Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented.

  1. Structural analysis and modeling of water reactor fuel rod behavior

    International Nuclear Information System (INIS)

    An important aspect of the design and analysis of nuclear reactor is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system under normal and emergency operating conditions. To achieve these objectives and in order to provide a suitable computer code based on fundamental material properties for design and study of the thermal-mechanical behavior of water reactor fuel rods during their irradiation life and also to demonstrate the fuel rod design and modeling for students, The KIANA-1 computer program has been developed by the writer at Amir-Kabir university of technology with support of Atomic Energy Organization of Iran. KIANA-1 is an integral one-dimensional computer program for the thermal and mechanical analysis in order to predict fuel rods performance and also parameter study of Zircaloy-clad UO2 fuel rod during steady state conditions. The code has been designed for the following main objectives: To give a solution for the steady state heat conduction equation for fuel as a heat source and clad by using finite difference, control volume and semi-analytical methods in order to predict the temperature profile in the fuel and cladding. To predict the inner gas pressures due to the filling gases and released gaseous fission products. To predict the fission gas production and release by using a simple diffusion model based on the Booth models and an empirical model. To calculate the fuel-clad gap conductance for cracked fuel with partial contact zones to a closed gap with strong contact. To predict the distribution of stress in three principal directions in the fuel and sheet by assuming one-dimensional plane strain and asymmetric idealization. To calculate the strain distribution in three principal directions and the corresponding deformation in the fuel and cladding. For this purpose the permanent strain such as creep or plasticity as well as the thermoelastic deformation and also the swelling, densification, cracking

  2. Experimental fuel rod stored energy determination. STEED I project

    Energy Technology Data Exchange (ETDEWEB)

    Engman, U.; Malen, K. [Studsvik Nuclear AB, Nykoeping (Sweden)

    1999-06-01

    The objective of the STEED I (STored Energy/Enthalpy Determination) project was to evaluate an experimental method for producing accurate and reliable data concerning the stored energy in fuel rods during operation. The STEED data should provide useful information for LOCA evaluation, fuel design and thermo-mechanical modelling. Stored energy refers to the amount of heat, which at a certain time is stored within the fuel. Physical properties of the fuel that affect the quantity of stored energy are radial power profile, burnup, fuel geometry, fuel density and thermal conductivity and heat capacity of the fuel pellet, and the gas gap conductance. The quantity of stored energy is conveniently studied under transient conditions when all, or part of the stored heat is released. This work describes determination of the stored energy by evaluating scram tests. The R2 test reactor is well suited for this type of experiments, where the thermal response of different types of fuel rods can be evaluated and compared. Scrams have been performed with the intent to evaluate the fuel rod stored energy before the scram. Methods have been developed for evaluation of the stored energy from the scram response It was found that the time dependence for a large part of the heat release from the rod could be described by a single time constant. Evaluations of the time constant have been made from the data in different ways. The stored energy has been evaluated integrating the exponential decay. The integral of the exponential decay is the initial power multiplied by the time constant. This means that differences in the stored energy due to, for instance, rod properties or rod power dependence are best studied using the same time constant. The scram response was modelled with the TOODEE2 transient code. The calculations gave a time constant of about 4 s and very little power dependence. The experimental result is a time constant around 4 s. The small differences in the measurement results

  3. Fuel element reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    The PUSPATI TRIGA Reactor has been utilized for more than 25 years using the same fuel elements and control rods. Generally, there are four control rods being used to control the neutron production inside the reactor core. A maintenance program has been developed to ensure its integrity, capability and safety of the reactor and it has been maintained twice a year since the first operation in 1982. The activities involve during the maintenance period including fuel elements and control rods inspections, electronics and mechanical systems, and others related works. During the maintenance in August 2008, there are some irregularities found on the fuel follower control rods and needed to be replaced. Even though the irregularities was not contributed into any unwanted incident, it were decided to replace with new control rods to avoid any potential hazards and unsafe condition occurred during operation later. Replacing any of the control rods would involved in imbalance of neutron flux and power distribution inside the core. Therefore, a number of fuel elements need to be reshuffled in order to compensate the neutron flux and power distribution as well as to balance the fuel elements burn-up in the core. This paper will described the fuel elements reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA Reactor. (Author)

  4. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  5. Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

    International Nuclear Information System (INIS)

    An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

  6. Test requirement for PIE of HANARO irradiated fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Lim, I. C.; Cho, Y. G

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U{sub 3}Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE.

  7. Optimization of fuel reloads for a BWR using the ant colony system; Optimizacion de recargas de combustible para un BWR usando el sistema de colonia de hormigas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J. [Universidad Autonoma del Estado de Mexico, Facultad de Ingenieria, Cerro de Coatepec s/n, Ciudad Universitaria, 50110 Toluca, Estado de Mexico (Mexico); Ortiz S, J. J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jaime.es.jaime@gmail.com

    2009-10-15

    In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

  8. Experimental investigation of the enthalpy- and mass flow-distribution in 16-rod clusters with BWR-PWR-geometries and conditions

    International Nuclear Information System (INIS)

    The enthalpy- and mass-flow-distribution at the outlet of two different 16-rod cluster test sections with uniform heating in axial and radial direction under steady state conditions has been measured for the first time by simultaneous sampling of 5 from 6 present characteristic subchannels in the bundle using the isokinetic technique and analysing the outlet quantities by a calorimetic method. The test-sections are provided with typical geometrical configurations for BWR s (70 bars; test section PELCO-S) and PWR s (160 bars; test-section EUROP). The latter has also been tested under BWR conditions (70 bars) to study the influence of geometry and pressure. The results showed the abnormal behaviour of the corner subchannel under BWR typical conditions (70 bars) which could not be found for PWR conditions (160 bars) and which is only an effect of pressure and not of geometry. The analysis of the experimental data confirms the usefullness of the subchannel sampling technique for the better understanding of the complex thermohydraulic phenomena under two-phase flow conditions in multirod bundles. Calculations of subchannel resistance coefficients for both types of spacers under one-phase flow conditions have been made with a special sub-structure method which showed a rather high local value of the corner subchannel. With the local drag coefficents the total resistance of the spacer has been evaluated and agreed well with measured values under adiabatic conditions. The measured subchannel data permit a direct valuation and examination of respective computer codes in a fundamental manner which are, however, not subject of this report

  9. FREC-3: a computer program to analyze stress and strain of fuel rods in accordance with fuel-rod irradiation history

    International Nuclear Information System (INIS)

    FREC-3 (Fuel Realiability Evaluation Code Version-3) analyzes the changes of fuel-rod design parameters in accordance with irradiation history, and evaluates the cladding strain, important fuel rods safety. The program is based on reports of the CYGRO developed by Westinghouse Electric Corporation (U.S.). This report describes the calculation procedure and models in FREC-3. (author)

  10. Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

    Science.gov (United States)

    Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; Nadiga, Balasubramanya T.; Karoutas, Zeses; Berndt, Markus

    2016-10-01

    Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuel rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid-structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.

  11. Criticality calculations for a spent fuel storage pool for a BWR type reactor

    International Nuclear Information System (INIS)

    In this work, the methodology for the calculation of the constant of effective multiplication for the arrangement of spent fuel assemblies in the pool of a BWR type reactor is shown. Calculations were done for the pool of spent fuel specified in FSAR and for the assemblies that is thought a conservative composition of high enrichment and without Gadolinium, giving credit to the stainless steel boxes of the frames that keep the assemblies. To carry out this simulation, RECORD and MIXQUIC codes were used. With record code, macroscopic cross sections, two energy groups, for the characteristics of the thought assemblies were obtained. Cross sections, as well as the dimensions of the frames that keep the fuel assemblies were used as input data for MIXQUIC code. With this code, criticality calculations in two dimensions were done, supposing that there is not leak of neutrons along the axial of the main line. Additional calculations, supposing changes in the temperature, distance among fuel assemblies and the thickness of the stainless steel box of the frame were done. The obtained results, including the effect in tolerances due to temperature, weight and thickness, show that the arrangement in the pool, when frames are fully charged, is subcritical by less than 5% in δK. (Author)

  12. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Warmann, Stephan A. [Portage, Inc., Idaho Falls, ID (United States); Rusch, Chris [NAC International, Inc., Norcross, GA (United States)

    2014-03-01

    , inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage. To document the initial condition of the used fuel prior to emplacement in a storage system, “sister ” fuel rods will be harvested and sent to a national laboratory for characterization and archival purposes. This report supports the demonstration by describing how sister rods will be shipped and received at a national laboratory, and recommending basic nondestructive and destructive analyses to assure the fuel rods are adequately characterized for UFDC work. For this report, a hub-and-spoke model is proposed, with one location serving as the hub for fuel rod receipt and characterization. In this model, fuel and/or clad would be sent to other locations when capabilities at the hub were inadequate or nonexistent. This model has been proposed to reduce DOE-NE’s obligation for waste cleanup and decontamination of equipment.

  13. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    International Nuclear Information System (INIS)

    , inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage. To document the initial condition of the used fuel prior to emplacement in a storage system, ''sister'' fuel rods will be harvested and sent to a national laboratory for characterization and archival purposes. This report supports the demonstration by describing how sister rods will be shipped and received at a national laboratory, and recommending basic nondestructive and destructive analyses to assure the fuel rods are adequately characterized for UFDC work. For this report, a hub-and-spoke model is proposed, with one location serving as the hub for fuel rod receipt and characterization. In this model, fuel and/or clad would be sent to other locations when capabilities at the hub were inadequate or nonexistent. This model has been proposed to reduce DOE-NE's obligation for waste cleanup and decontamination of equipment.

  14. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B4C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

  15. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    International Nuclear Information System (INIS)

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

  16. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    Energy Technology Data Exchange (ETDEWEB)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  17. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Gonzalez, Jesus S. [Univ. Politecnica de Madrid (Spain); Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William BJ J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

  18. Experimental simulation of LOCA in a PWHR : analytical study of similarity of thermal response between fuel rod simulators and nuclear fuel rods under reflood conditions

    International Nuclear Information System (INIS)

    For the safety analysis of a nuclear reactor, the similarity of the thermal response of an electrically heated Fuel Rod Simulator (FRS), mostly used in Loss-of-Coolant-Accident (LOCA) experiments, to that of a nuclear fuel rod is of great significance. The present analysis describes the characteristics and the similarity of thermal response fuel rods under reflood conditions of LOCA. The analysis has shown that the thermal response of a nuclear fuel rod can be well simulated by the use of an electrically heated FRS. (author). 7 refs., 12 figs

  19. Drilling Experiments of Dummy Fuel Rods Using a Mock-up Drilling Device and Detail Design of Device for Drilling of Irradiated Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Yong; Lee, H. K.; Chun, Y. B.; Park, S. J.; Kim, B. G

    2007-07-15

    KAERI are developing the safety evaluation method and the analysis technology for high burn-up nuclear fuel rod that is the project, re-irradiation for re-instrumented fuel rod. That project includes insertion of a thermocouple in the center hole of PWR nuclear fuel rod with standard burn-up, 3,500{approx}4,000MWD/tU and then inspection of the nuclear fuel rod's heat performance during re-irradiation. To re-fabricate fuel rod, two devices are needed such as a drilling machine and a welding machine. The drilling machine performs grinding a center hole, 2.5 mm in diameter and 50 mm in depth, for inserting a thermocouple. And the welding machine is used to fasten a end plug on a fuel rod. Because these two equipment handle irradiated fuel rods, they are operated in hot cell blocked radioactive rays. Before inserting any device into hot cell, many tests with that machine have to be conducted. This report shows preliminary experiments for drilling a center hole on dummy of fuel rods and optimized drilling parameters to lessen operation time and damage of diamond dills. And the design method of a drilling machine for irradiated nuclear fuel rods and detail design drawings are attached.

  20. Study on flow-induced vibration of the fuel rod in HTTR

    International Nuclear Information System (INIS)

    This study was performed in order to investigate flow-induced vibration characteristics of a fuel rod in HTTR (High Temperature engineering Test Reactor) from both an experiment and a numerical simulation. Two kinds of fuel rods were used in this experiment: one was a graphite rod which simulated a specification of the HTTR's fuel rod and the other was an aluminum rod whose weight was a half of the graphite one. The experiment was carried out up to Re = 31000 using air at room temperature and pressure. Air flowed downstream in an annular passage which consisted of the fuel rod and the graphite channel. Numerical simulations by fluid and frequency equations were also carried out. Numerical and experimental results were then compared. The following conclusions were drived: (1) The fuel rod amplitudes increase with the flow rate and with a decrease of the fuel rod weight. (2) The fuel rod amplitudes are obtained by δ/De = 2.22 x 10-10Re1.43, 9000 ≤ Re ≤ 31000, where δ is a vibration amplitude, De is a hydraulic diameter and Reis Reynolds number. (3) The fuel rod frequencies shift from lower natural frequency to higher as the flow rate increases. (4) The flow-induced vibration behavior of the fuel rod can simulate well by simultaneous equations which used the turbulence model for fluid and the mass model for vibration of the fuel rod. (author)

  1. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  2. Licensing of spent fuel dry storage and consolidated rod storage

    International Nuclear Information System (INIS)

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs

  3. Pressure equalization system in PWR-fuel rods

    International Nuclear Information System (INIS)

    The pressure equalization system, developed on the basis of activated charcoal, is capable of reducing the internal pressure rise in fuel rods by adsorption of the fission gases. He-prepressure does not affect the system and Helium will not be adsorbed. Irradiation does not reduce the adsorption capacity of activated charcoal down to an unacceptable limit. Shaped activated charcoal is a suitable material which can be well defined and characterized. Feasible techniques of activating and assembling methods can be proposed. (orig.)

  4. Welding of zircaloy tubes for fuel rods fabrication

    International Nuclear Information System (INIS)

    The TIG process form welding zircaloy tube-end plug to fabricate UO2 pellet fuel rods is presented. The zircaloy tube is used as weld by welding zircaloy end-plugs inserted into ends. The technical aspects of this welding process, necessary equipment, and weld properties are discussed. The main problems involved with this type of weld and the nondestructive and destructive weld quality controls are emphasized. (Author)

  5. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    OpenAIRE

    Alroumi Fawaz; Kim Donghoon; Schow Ryan; Jevremovic Tatjana

    2016-01-01

    Control rod reactivity (worths) for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I) are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is ...

  6. Preliminary design report for the prototypical fuel rod consolidation system

    International Nuclear Information System (INIS)

    This report documents NUTECH's preliminary design of a dry, spent fuel rod consolidation system. This preliminary design is the result of Phase I of a planned four phase project. The present report on this project provides a considerable amount of detail for a preliminary design effort. The design and all of its details are described in this Preliminary Design Report (PDR). The NUTECH dry rod consolidation system described herein is remotely operated. It provides for automatic operation, but with operator hold points between key steps in the process. The operator has the ability to switch to a manual operation mode at any point in the process. The system is directed by the operator using an executive computer which controls and coordinates the operation of the in-cell equipment. The operator monitors the process using an in-cell closed circuit television (CCTV) system with audio output and equipment status displays on the computer monitor. The in-cell mechanical equipment consists of the following: (1) two overhead cranes with manipulators; (2) a multi-degree of freedom fuel handling table and its clamping equipment; (3) a fuel assembly end fitting removal station and its tools; (4) a consolidator (which pulls rods, assembles the consolidated bundle and loads the canister); (5) a canister end cap welder and weld inspection system; (6) decontamination systems; and (7) the CCTV and microphone systems

  7. Technique of manufacturing specimen of irradiated fuel rods

    International Nuclear Information System (INIS)

    Technique of manufacturing specimen of irradiated fuel rods to perform efficient PIE is developed by analyzing the relation between requiring time of manufacturing specimen and manufacturing method in irradiated fuel rods. It takes within an hour to grind 1 mm of specimen thickness under 150 rpm in speed of grinding, 600 g gravity in force using no.120, no.240, no.320 of grinding paper. In case of no.400 of grinding paper, it takes more an hour to grind the same thickness as above. It takes up to a quarter to grind 80-130 μm in specimen thickness using no.400 of grinding paper. When grinding time goes beyond 15 minutes, the grinding thickness of specimen does not exist. The polishing of specimen with 150 Rpms in speed of grinding machine, 600 g gravity in force, 10 minutes in polishing time using diamond paste 15 μm on polishing cloths amounts to 50 μm in specimen thickness. In case of diamond paste 9 μm on polishing cloth, the polishing of specimen amounts to 20 μm. The polishing thickness of specimen with 15 minutes in polishing time using 6 μm, 3 μm, 1 μm, 1/4 μm does not exist. Technique of manufacturing specimen of irradiated fuel rods will have application to the destructive examination of PIE. (author). 6 refs., 1 tab., 10 figs

  8. Simulation of fuel rod irradiation capsules in water loops by electric heater rods

    International Nuclear Information System (INIS)

    The out of pile simulation of irradiation devices was carried out by J.E.N. in the frame of the KfK-JEN joint experiment for irradiation of fast reactor fuel rods (IVO-FR2-Vg7). A typical single-wall-Nak (22% Na, 78% K) electrical heated capsule was fabricated and hydraulical tests were done. The capsule was instrumented with 10 thermocouples in order to obtain the radial temperature profile into the capsule in function of the electrical rod power (max. 215 w/cm), flow rate (max. 2,4 m3/h) and coolant temperature (max. 60degC). The experimental values are compared to the Tecap-Code results. (author)

  9. Criteria for removal of defective fuel rod from fuel assembly under repair without cladding rupture

    International Nuclear Information System (INIS)

    During repair of a failed fuel assembly (FA) there is a risk of cladding rupture while a defective fuel rod is forced out of the assembly skeleton. To reduce the corresponding risks, a program of experimental and analytical studies for WWER fuel was performed. It resulted in formulation of criteria for successful removal of the defective fuel rod from the FA under repair. 'Successful' means that no cladding rupture occurs. The paper summarizes the available data of post-irradiation examinations of WWER FAs with leaking fuel rods. A technique for express estimation of hydrogen content in cladding of a defective fuel rod is presented. The degradation of cladding mechanical properties can be estimated with this technique as well. A criterion of severe secondary hydriding involved in the risk analysis is also discussed. Finally, it is shown how the information on operation conditions may be used for prompt evaluation of the limiting force for successful removal of a defective fuel rod during FA repair in the inspection stand. (author)

  10. CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

    Directory of Open Access Journals (Sweden)

    Dávid Halabuk

    2015-12-01

    Full Text Available The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

  11. Design of a fuel recharge for a BWR using advanced optimization systems

    International Nuclear Information System (INIS)

    The fuel recharge design for a BWR reactor it was carried out, which includes the design of four fuel cells to form an assembly, the accommodation design of fresh and partially consumed assemblies and the control bars pattern design to use along an operation cycle. The three stages were approached as optimization problems using different computational tools, each one of those includes an objective function to measure quantitatively the evolution of the different candidate solutions. With the tool used in the fuel cells design that makes use of the tabu search technique its were obtained cells that showed to be lightly more reactive that other similar taken as reference. With the four designed cells it was formed a fuel assembly that turn out to have an average enrichment lightly smaller to the one of another assembly similar taken as reference. In the recharge pattern design it was used another optimization tool, also based on tabu search to obtain the accommodation of 108 fresh fuels and 336 partially consumed, fulfilling the conditions imposed to operate with the core strategies with control cells (CCC) and of low leakage. A Haling calculation reported that with the obtained accommodation it was achievement to increase in 8% the cycle length with regard to the one obtained using a similar reference pattern. In the design of the control bars patterns it was used a tool based on the use of the genetic algorithms to obtain the placement patterns of the control bars along an operation cycle. The search tool only uses the bars of the A2 sequence and it makes use of the 1/8 symmetry of the core, with that the number of used control bars it decreases at 5. Also the use of control bars in intermediate positions is also avoided. With the obtained patterns a cycle length is obtained that is lightly bigger to the reported value in a Haling calculation. (Author)

  12. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  13. Optimization of fuel cells for BWR based in Tabu modified search

    International Nuclear Information System (INIS)

    The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

  14. The ‘Fuel Rod Analysis ToolBox’: A general program for preparing the input of a fuel rod performance code

    International Nuclear Information System (INIS)

    Highlights: • An input-file preparation tool for fuel rod performance codes is developed • The new tool can be used for benchmarks of fuel rod performance codes • The new tool can condense irradiation histories, synchronise and merge signals • The new toolbox for fuel rod analysis has a graphical user interface • The new toolbox for fuel rod analysis is available for free - Abstract: This paper gives an overview of the new ‘Fuel Rod Analysis Toolbox’, which is a program for the pre-processing of input data for fuel rod performance codes with a graphical user interface. It consists of three different modules that can handle several tasks such as data condensation, merging and synchronization. The ‘Fuel Rod Analysis Toolbox’ helps: • to reduce the amount of input data, • to simplify the setup of input files for complex data sets with input from experimental data as well as input resulting from neutronics or thermo-hydraulics codes, and • to reduce computation time. These advantages are already evident for complex fuel rod analyses employing a conventional one-and-a-half-dimensional code but they become even more important for two- or three-dimensional approaches

  15. The analysis of failed nuclear fuel rods by gamma computed tomography

    Science.gov (United States)

    Dobrin, Relu; Craciunescu, Teddy; Tuturici, Ioan Liviu

    1997-07-01

    The failure of the cladding of an irradiated nuclear fuel rod can lead to the loss of some γ-radioactive fission products. Consequently the distribution of these fission products is altered in the cross-section of the fuel rod. The modification of the distribution, obtained by gamma computed tomography, is used to determine the integrity of the fuel cladding. The paper reports an experimental result, obtained for a CANDU-type fuel rod, irradiated in a TRIGA 14 MWth reactor.

  16. BWR feedwater nozzle and control rod drive return line nozzle cracking: resolution of generic technical activity A-10. Technical report

    International Nuclear Information System (INIS)

    This report summarizes work performed by the NRC staff in the resolution of Generic Technical Activity A-10, 'BWR Nozzle Cracking'. Generic Technical Activity A-10 is one of the generic technical subjects designated as 'unresolved safety issues' pursuant to Section 210 of the Energy Reorganization Act of 1974. The report describes the technical issues, the technical studies and analyses performed by the General Electric Company and the NRC staff, the staff's technical positions based on these studies, and the staff's plans for continued implementation of its technical positions. It also provides information for further work to resolve the non-destructive examination issue

  17. Multidimensional simulations of hydrides during fuel rod lifecycle

    Science.gov (United States)

    Stafford, D. S.

    2015-11-01

    In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim.

  18. A basic design of a double cladding fuel rod to control the irradiation temperature on nuclear fuels

    International Nuclear Information System (INIS)

    An instrumented capsule for a nuclear fuel irradiation test (hereinafter referred to as 'instrumented fuel capsule') has been developed to measure fuel characteristics, such as a fuel center and surface temperature, the internal pressure of a fuel rod, a fuel pellet elongation and neutron flux, during an irradiation test at HANARO. And six types of dual instrumented fuel rods, which allow for two characteristics to be measured simultaneously in one fuel rod, have been developed to enhance the efficiency of an irradiation test using an instrumented fuel capsule at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technologies for high temperature nuclear fuel irradiation tests at HANARO. The purpose of this paper is to control the temperature of nuclear fuels during an irradiation test at HANARO. Therefore we basically designed a double cladding fuel rod and an instrumented fuel capsule basically. The basic design of a double cladding rod was based on out-pile tests using mockups and the thermal analyses using some relevant codes. This paper presents the design and fabrication of the double cladding fuel rod mockups, the results of the out-pile tests, the results of the temperature calculation and the basic design of a double cladding fuel rod and an instrumented fuel capsule. (author)

  19. A comparison of thermal algorithms of fuel rod performance code systems

    International Nuclear Information System (INIS)

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

  20. Joining techniques for core flow test loop fuel rod simulators

    International Nuclear Information System (INIS)

    Laser welding and furnace brazing techniques have been developed to join subassemblies for fuel rod simulators (FRSs) that have survived up to 1000-h steady-state operation at 700 to 11000C cladding temperatures and over 5000 thermal transients, ranging from 10 to 500C/s. A pulsed-laser welding procedure uses small diameter filler wire to join one end of a resistance heating element to a tubular conductor. The other end of the heating element is laser welded to an end plug, which in turn is welded to a central conductor. Before these welding operations, the intermediate material conductors (either tubular or rod) are vacuum brazed to matching copper leads. On room temperature tensile testing, 10 of 11 brazements between copper and nickel rods failed in the copper rather than the brazement. The thin walls and ductility of the copper and nickel tubular conductors caused joint machining and fitup problems. Accordingly, it has not been possible to consistently produce tensile test samples of brazed dissimilar metal tubular conductors that will fail outside the joint area. A unique tubular electrode carrier has also been developed for gas tungsten arc welding FRSs to the tubesheet of a test assembly. Two seven-rod mockups of the simulator-to-tubesheet joint area were welded and successfully cycled 500 times from 3700C (6980F) down to 1000C (2120F) with an internal pressure of 11.72 MPa (1700 psi). No leakage was detected by helium mass spectrometry, either before or after testing. Modified versions of the electrode carrier were developed for brazing electrical leads to the upper ends of the FRSs. Satisfactory brazes have been made on both single-rod mockups and arrays of simulators

  1. Research on Integrity of High Burnup Spent Fuel Under the Long Term Dry Storage

    International Nuclear Information System (INIS)

    Objectives were to acquire the following behaviour data by dynamic load impact tests on high burnup spent fuel rods of BWR and PWR and to improve the guidance of regulation of spent fuel storage and transportation. (1) The limit of load and strain for high burnup fuel in the cask drop accident. (2) The amount of deformation of high burnup fuel rods under dynamic load impact. (3) The amount of fuel pellet material released from fuel rods under dynamic load impact

  2. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  3. Thermomechanical evaluation of BWR fuel elements for procedures of preconditioned with FEMAXI-V

    International Nuclear Information System (INIS)

    The limitations in the burnt of the nuclear fuel usually are fixed by the one limit in the efforts to that undergo them the components of a nuclear fuel assembly. The limits defined its provide the direction to the fuel designer to reduce to the minimum the fuel failure during the operation, and they also prevent against some thermomechanical phenomena that could happen during the evolution of transitory events. Particularly, a limit value of LHGR is fixed to consider those physical phenomena that could lead to the interaction of the pellet-shirt (Pellet Cladding Interaction, PCI). This limit value it is related directly with an PCI limit that can be fixed based on experimental tests of power ramps. This way, to avoid to violate the PCI limit, the conditioning procedures of the fuel are still required for fuel elements with and without barrier. Those simulation procedures of the power ramp are carried out for the reactor operator during the starting maneuvers or of power increase like preventive measure of possible consequences in the thermomechanical behavior of the fuel. In this work, the thermomechanical behavior of two different types of fuel rods of the boiling water reactor is analyzed during the pursuit of the procedures of fuel preconditioning. Five diverse preconditioning calculations were carried out, each one with three diverse linear ramps of power increments. The starting point of the ramps was taken of the data of the cycle 8 of the unit 1 of the Laguna Verde Nucleo electric Central. The superior limit superior of the ramps it was the threshold of the lineal power in which a fuel failure could be presented by PCI, in function of the fuel burnt. The analysis was carried out with the FEMAXI-V code. (Author)

  4. Implement of MOX fuel assemblies in the design of the fuel reload for a BWR; Implemento de ensambles de combustible MOX en el diseno de la recarga de combustible para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Alonso V, G.; Palacios H, J. C., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    At the present time the use of mixed oxides as nuclear fuel is a technology that has been implemented in mixed reloads of fuel for light water reactors. Due to the plutonium production in power reactors, is necessary to realize a study that presents the plutonium use like nuclear fuel. In this work a study is presented that has been carried out on the design of a fuel assembly with MOX to be proposed in the supply of a fuel reload. The fissile relationship of uranium to plutonium is presented for the design of the MOX assembly starting from plutonium recovered in the reprocessing of spent fuel and the comparison of the behavior of the infinite multiplication factor is presented and of the local power peak factor, parameters of great importance in the fuel assemblies design. The study object is a fuel assembly 10 x 10 GNF2 type for a boiling water reactor. The design of the fuel reload pattern giving fuel assemblies with MOX, so the comparison of the behavior of the stop margin for a fuel reload with UO{sub 2} and a mixed reload, implementing 12 and 16 fuel assemblies with MOX are presented. The results show that the implement of fuel assemblies with MOX in a BWR is possible, but this type of fuels creates new problems that are necessary to study with more detail. In the development of this work the calculus tools were the codes: INTREPIN-3, CASMO-4, CMSLINK and SIMULATE-3. (Author)

  5. Method for verifying the pressure in a nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Disclosed is a method of accurately verifying the pressure contained in a sealed pressurized fuel rod by utilizing a pressure balance measurement technique wherein an end of the fuel rod extends through and is sealed in a wall of a small chamber. The chamber is pressurized to the nominal (desired) fuel rod pressure and the fuel rod is then pierced to interconnect the chamber and fuel rod. The deviation of chamber pressure is noted. The final combined pressure of the fuel rod and drill chamber is substantially equal to the nominal rod pressure; departure of the combined pressure from nominal is in direct proportion to departure of rod pressure from nominal. The maximum error in computing the rod pressure from the deviation of the combined pressure from nominal is estimated at plus or minus 3.0 psig for rod pressures within the specified production limits. If the rod pressure is corrected for rod void volume using a digital printer data record, the accuracy improves to about plus or minus 2.0 psig

  6. VIPRE-W / MEFISTO-T - A mechanistic tool for transient prediction of dryout in BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Adamsson, C., E-mail: carl.adamsson@psi.ch [Westinhouse Electric Sweden, Vasteras (Sweden); Paul Scherrer Institut, Villigen (Switzerland); Le Corre, J-M., E-mail: lecorrjm@westinghouse.com [Westinhouse Electric Sweden, Vasteras (Sweden)

    2011-07-01

    The VIPRE-W/MEFISTO-T code package constitutes a simplified approach to sub-channel film-flow analysis whereby the transport equations for the liquid films are decoupled from each other. The approach allows fast and robust simulation with high axial resolution of realistic BWR transients. It has previously been shown that a steady-state version of the model agrees well with dryout measurements in full-scale fuel assembly mock-ups performed at the Westinghouse FRIGG loop. In this paper, we present validation of the transient version of the code with around 300 transient dryout experiments from the same loop. The transients involve realistic variations of flow and power and three different axial power distributions at conditions typical for BWR operation. The results from the film-flow analysis show high precision in the dryout prediction but a hitherto unexplained bias that reduces the accuracy. (author)

  7. Performance evaluation of the Loviisa advanced type fuel rods

    International Nuclear Information System (INIS)

    The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

  8. Visualization test facility of nuclear fuel rod emergency cooling system

    International Nuclear Information System (INIS)

    The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

  9. Analysis of the structural integrity of the fuel rod cladding based on ring compression tests

    International Nuclear Information System (INIS)

    Due to the reduced amount of material involved and the relatively simple test set-up, Ring Compression Tests (RCT) on fuel rod cladding specimens has become a well-accepted test to determine the conditions resulting in a brittle response on the cladding. Indeed, from its application under LOCA conditions, also it is used under the Spent Fuel Storage and Transportation conditions. Although the RCT may run the involved material through three stages: elastic, elasto-plastic and damage propagation and relevant information on material properties may be obtained, the non-homogenous stress and strain conditions makes the analysis of the test results, difficult. Even though, some efforts have successfully provided key cladding performance parameters such as the fracture toughness. Others approaches use the RCT as a screening test to determine conditions resulting in a Ductile-to-Brittle transition based on a selected criterion. This paper proposes a criterion from the RCT results based on first principles to address cladding ductility under the pinch loads that occurs during the transportation accident of the cask horizontal drop. The insights gained from a mechanical analysis of the RCT are applied on a number of RCT performed on unirradiated pre-hydrided specimens. Besides, RCT results performed on BWR irradiated cladding with several degrees of radial reorientation of the hydrides, imposed by a previous creep test, are also analyzed following the same approach. Based on this analysis and the expected diametric displacement, allowed by the end of irradiation pellet to clad gap and the outward cladding creep during drying and storage in a dry cask, a criterion is determined. (author)

  10. Non-destructive Testing Dummy Nuclear Fuel Rods by Neutron Radiography

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng

    2013-01-01

    As a unique non-destructive testing technique,neutron radiography can be used to measure nuclear fuel rods with radioactivity.The images of the dummy nuclear fuel rods were obtained at the CARR.Through imaging analysis methods,the structure defections,the hydrogen accumulation in the cladding and the 235U enrichment of the pellet were studied and analyzed.Experiences for non-destructive testing real PWR nuclear fuel rods by NR

  11. FDD-1 System On-line Monitoring Fuel Rod Failure of Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; JISong-tao; GAOYong-guang; YINZhen-guo; HANChuan-bin

    2003-01-01

    The FDD-1 system developed by CIAE for on-line monitoring fuel rod failure of nuclear power plant consists of γ-ray detector, γ-ray spectrum analyzer, computer, and an analysis code for evaluating the status of fuel rod failure. It would be determined that the fuel rod failure occurs when a large amount of γ activity increases in the primary system measured by γ-ray detector near the CVCS.

  12. Optimization of fuel cells for BWR based in Tabu modified search; Optimizacion de celdas de combustible para BWR basada en busqueda Tabu modificada

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Francois L, J.L. [Facultad de Ingenieria, UNAM, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Palomera P, M.A. [Facultad de Ingenieria, UNAM, Posgrado en Ingenieria en Computacion, Circuito exterior s/n, Ciudad Universitaria, Mexico, D.F. (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2004-07-01

    The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

  13. BWR-spent fuel transport and storage with the TN trademark 9/4 and TN trademark 24BH casks

    International Nuclear Information System (INIS)

    The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks in ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., as Muehleberg Nuclear Power Plant owner, is involved in this process and has selected to store its spent fuel, a new high capacity dual-purpose cask, the TN trademark 24BH. For the transport in a medium size cask, COGEMA LOGISTICS has developed a new cask, the TN trademark 9/4, to replace the NTL9 cask, which performed numerous transports of BWR spent fuel in the past decades. Licensed IAEA 1996, the TN trademark 9/4 is a 40 ton transport cask, for 7 BWR high burn-up spent fuel assemblies. The spent fuel assemblies can be transferred in the ZWILAG hot cell in the TN trademark 24BH cask. The first use of these casks took place in 2003. Ten TN trademark 9/4 transports were performed, and one TN trademark 24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN trademark 24BH high capacity dual purpose cask, the TN trademark 9/4 transport cask and describe in detail their characteristics and possibilities

  14. Flow induced vibration forces on a fuel rod by LES CFD analysis

    International Nuclear Information System (INIS)

    The purpose of the present study is to evaluate the feasibility of use of CFD Large Eddy Simulation (LES) modeling techniques in CD-adapco CFD code STAR-CCM+ to calculate the instantaneous stress tensor on the fuel rod wall and then utilize these data for mechanical calculations. Transient hydraulic forces on the fuel rod resulting from the CFD model are linked to the Westinghouse VITRAN code to predict fuel rod vibration response. The coupled CFD/mechanical solution has provided a reasonable prediction of fuel rod vibration and a more accurate representation of all the important physics and excitation forces. (author)

  15. Re-fabrication and re-instrumentation of irradiated LWR fuel rods for irradiation testing at the HFR Petten

    International Nuclear Information System (INIS)

    LWR fuel testing at the High Flux Reactor (HFR) and the Hot Cells at Petten has been successfully performed with pre-irradiated fuel rod segments. The testing methods have been extended with hot cell techniques for re-fabrication of test fuel rods from full length fuel rods from power reactors; re-instrumentation of pre-irradiated fuel rod segments with pressure sensors; and instrumentation of re-fabricated fuel rods or fuel rod segments with central thermocouple and/or pressure sensors. 5 refs, 9 figs, 5 tabs

  16. A methodology for obtaining the control rods patterns in a BWR using systems based on ants colonies

    International Nuclear Information System (INIS)

    In this work the AZCATL-PBC system based on a technique of ants colonies for the search of control rods patterns of those reactors of the Nuclear Power station of Laguna Verde (CNLV) is presented. The technique was applied to a transition cycle and one of balance. For both cycles they were compared the kef values obtained with a Haling calculation and the control rods pattern proposed by AZCATL-PBC for a burnt one fixed. It was found that the methodology is able to extend the length of the cycle with respect to the Haling prediction, maintaining sure to the reactor. (Author)

  17. Development for analysis system of rods enrichment of nuclear fuels

    International Nuclear Information System (INIS)

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  18. Radial optimization of a BWR fuel cell using genetic algorithms; Optimizacion radial de una celda de combustible BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P. [UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2006-07-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U{sup 235} and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix

  19. SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

    International Nuclear Information System (INIS)

    Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.)

  20. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  1. IAMBUS, a computer code for the design and performance prediction of fast breeder fuel rods

    International Nuclear Information System (INIS)

    IAMBUS is a computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fast breeder fuel rods. The code deals with steady, non-steady and transient operating conditions and enables to predict in-pile behavior of fuel rods in power reactors as well as in experimental rigs. Great effort went into the development of a realistic account of non-steady fuel rod operating conditions. The main emphasis is placed on characterizing the mechanical interaction taking place between the cladding tube and the fuel as a result of contact pressure and friction forces, with due consideration of axial and radial crack configuration within the fuel as well as the gradual transition at the elastic/plastic interface in respect to fuel behavior. IAMBUS can be readily adapted to various fuel and cladding materials. The specific models and material correlations of the reference version deal with the actual in-pile behavior and physical properties of the KNK II and SNR 300 related fuel rod design, confirmed by comparison of the fuel performance model with post-irradiation data. The comparison comprises steady, non-steady and transient irradiation experiments within the German/Belgian fuel rod irradiation program. The code is further validated by comparison of model predictions with post-irradiation data of standard fuel and breeder rods of Phenix and PFR as well as selected LWR fuel rods in non-steady operating conditions

  2. Thermomechanical evaluation of BWR fuel elements for procedures of preconditioned with FEMAXI-V; Evaluacion termomecanica de elementos combustible BWR para procedimientos de preacondicionado con FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Lucatero, M.A.; Ortiz V, J. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2006-07-01

    The limitations in the burnt of the nuclear fuel usually are fixed by the one limit in the efforts to that undergo them the components of a nuclear fuel assembly. The limits defined its provide the direction to the fuel designer to reduce to the minimum the fuel failure during the operation, and they also prevent against some thermomechanical phenomena that could happen during the evolution of transitory events. Particularly, a limit value of LHGR is fixed to consider those physical phenomena that could lead to the interaction of the pellet-shirt (Pellet Cladding Interaction, PCI). This limit value it is related directly with an PCI limit that can be fixed based on experimental tests of power ramps. This way, to avoid to violate the PCI limit, the conditioning procedures of the fuel are still required for fuel elements with and without barrier. Those simulation procedures of the power ramp are carried out for the reactor operator during the starting maneuvers or of power increase like preventive measure of possible consequences in the thermomechanical behavior of the fuel. In this work, the thermomechanical behavior of two different types of fuel rods of the boiling water reactor is analyzed during the pursuit of the procedures of fuel preconditioning. Five diverse preconditioning calculations were carried out, each one with three diverse linear ramps of power increments. The starting point of the ramps was taken of the data of the cycle 8 of the unit 1 of the Laguna Verde Nucleo electric Central. The superior limit superior of the ramps it was the threshold of the lineal power in which a fuel failure could be presented by PCI, in function of the fuel burnt. The analysis was carried out with the FEMAXI-V code. (Author)

  3. Behavior of AgInCd absorber material in Zry/UO{sub 2} fuel rod simulator bundles tested at high temperatures in the CORA facility

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L.; Hagen, S.; Hofmann, P.; Schanz, G.

    2009-01-15

    The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program is to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations are tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The PWR-type assemblies usually consist of 25 rods with 16 electrically heated fuel rod simulators and nine unheated rods (full-pellet and absorber rods). Bundle CORA-5 contained one Ag/In/Cd - steel absorber rod whereas two absorber rods were used in CORA-12, CORA-15, and CORA-9. The larger bundle CORA-7 contained 5 absorber rods. CORA-12 was terminated by quenching with water from the bottom. In CORA-15 the heated and unheated rods were pressurized to achieve pronounced clad ballooning. Bundle CORA-9 was tested with a system pressure of 1.0 MPa instead of 0.22 MPa. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy (Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. Rod destruction started with the failure of the absorber rod cladding at about 1200 C, i.e. about 250 K below the melting regime of steel. Penetration of the steel cladding was presumably caused by a eutectic interaction between steel and the zircaloy guide tube. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO{sub 2} interaction

  4. Common cause failure analysis of hydraulic scram and control rod systems in the Swedish and Finnish BWR plants

    International Nuclear Information System (INIS)

    The main task of the project included the analysis of the operating experiences at the BWRs of ABB Atom design, comprising 9 units in Sweden and 2 in Finland. International experience and reference information were also surveyed. A reference application was done for the Barsebaeck plant. This pilot study covered all systems which contribute to the reactor shutdown, including also the actuation relays at the interface to the reactor protection system. The Common Load Model was used as the quantification method, which proved to be a practicable approach. This method provides a consistent handling of failure combinatorics and workable extension to evaluate localized dependence between adjacent control rod and drive assemblies (CRDAs). As part of this project, instructions of handbook style were prepared for the CCF analysis of high redundancy systems. The primary focus in the analysis of operating experience was placed on the scram valves and CRDAs. Due to the limited component population, the experiences for the scram valve constitute only a few single failures and some potential but none actual CCF event. These insights are compatible with the generic data for these valves. The experiences for the CRDAs include several single failures, and some actual and many potential CCF events of varying degree of functional impact. Special emphasis was placed to identify any multiple failure or degradation indicating that adjacent rods would be more vulnerable to failure, because such phenomena are far more critical for the scram function as compared to failure of randomly placed rods. 17 refs

  5. Free-hanging bow measurements of LWBR fuel rods (LWBR Development Program)

    International Nuclear Information System (INIS)

    Special inspection equipment was developed to ensure that the fuel rods for the Light Water Breeder Reactor met the required straightness criteria. The fuel rods were hung in a vertical position and the free-hanging shape was measured. These data were then used analytically to predict both the forces required to constrain the rods in a grid array and the resultant restrained shape. The development of a computerized system which was used for measuring the free-hanging bow of fuel rods used in the LWBR core is described in this paper

  6. Underwater Nuclear Fuel Disassembly and Rod Storage Process and Equipment Description. Volume II

    International Nuclear Information System (INIS)

    The process, equipment, and the demonstration of the Underwater Nuclear Fuel Disassembly and Rod Storage System are presented. The process was shown to be a viable means of increasing spent fuel pool storage density by taking apart fuel assemblies and storing the fuel rods in a denser fashion than in the original storage racks. The assembly's nonfuel-bearing waste is compacted and containerized. The report documents design criteria and analysis, fabrication, demonstration program results, and proposed enhancements to the system

  7. Prediction of the local power factor in BWR fuel cells by means of a multilayer neural network

    International Nuclear Information System (INIS)

    To the beginning of a new operation cycle in a BWR reactor the reactivity of this it increases by means of the introduction of fresh fuel, the one denominated reload fuel. The problem of the definition of the characteristics of this reload fuel represents a combinatory optimization problem that requires significantly a great quantity of CPU time for their determination. This situation has motivated to study the possibility to substitute the Helios code, the one which is used to generate the new cells of the reload fuel parameters, by an artificial neuronal network, with the purpose of predicting the parameters of the fuel reload cell of a BWR reactor. In this work the results of the one training of a multilayer neuronal net that can predict the local power factor (LPPF) in such fuel cells are presented. The prediction of the LPPF is carried out in those condition of beginning of the life of the cell (0.0 MWD/T, to 40% of holes in the one moderator, temperature of 793 K in the fuel and a moderator temperature of 560 K. The cells considered in the present study consist of an arrangement of 10x10 bars, of those which 92 contains U235, some of these bars also contain a concentration of Gd2O3 and 8 of them contain only water. The axial location inside the one assembles of recharge of these cells it is exactly up of the cells that contain natural uranium in the base of the reactor core. The training of the neuronal net is carried out by means of a retro-propagation algorithm that uses a space of training formed starting from previous evaluations of cells by means of the Helios code. They are also presented the results of the application of the neuronal net found for the prediction of the LPPF of some cells used in the real operation of the Unit One of the Laguna Verde Nuclear Power station. (Author)

  8. Performance analysis of VVER-type fuel rods with the STOFFEL-1 computer code

    International Nuclear Information System (INIS)

    The main features of the fuel rod performance modelling computer code STOFFEL-1 are described. Submodels of the code are briefly characterized, and some results of comparisons between model predictions and experiments are presented. Examples of modelling calculations are given for some thermo-mechanical values of VVER-1000 fuel rods. (author)

  9. Automatic system of welding for nuclear fuel rods

    International Nuclear Information System (INIS)

    The welding process of nuclear fuel must be realized in an inert gas environment (He) and constant flow of this. In order to reach these conditions it is necessary to do vacuum at the chamber and after it is pressurized with the noble gas (purge) twice in the welding chamber. The purge eliminates impurities that can provoke oxidation in the weld. Once the conditions for initiating the welding are gotten, it is necessary to draw a graph of the flow parameters, pressure, voltage and arc current and to analyse those conditions in which have been carried out the weld. The rod weld must be free of possible pores or cracks which could provoke rod leaks, so reducing the probability of these failures should intervene mechanical and metallurgical factors. Automatizing the process it allows to do reliable welding assuring that conditions have been performed, reaching a high quality welding. Visually it can be observed the welding process by means of a mimic which represents the welding system. There are the parameters acquired such as voltage, current, pressure and flow during the welding arc to be analysed later. (Author)

  10. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  11. A Multi-Stage Wear Model for Grid-to-Rod Fretting of Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Blau, Peter Julian [ORNL

    2014-01-01

    The wear of fuel rod cladding against the supporting structures in the cores of pressurized water nuclear reactors (PWRs) is an important and potentially costly tribological issue. Grid-to-rod fretting (GTRF), as it is known, involves not only time-varying contact conditions, but also elevated temperatures, flowing hot water, aqueous tribo-corrosion, and the embrittling effects of neutron fluences. The multi-stage, closed-form analytical model described in this paper relies on published out-of-reactor wear and corrosion data and a set of simplifying assumptions to portray the conversion of frictional work into wear depth. The cladding material of interest is a zirconium-based alloy called Zircaloy-4, and the grid support is made of a harder and more wear-resistant material. Focus is on the wear of the cladding. The model involves an incubation stage, a surface oxide wear stage, and a base alloy wear stage. The wear coefficient, which is a measure of the efficiency of conversion of frictional work into wear damage, can change to reflect the evolving metallurgical condition of the alloy. Wear coefficients for Zircaloy-4 and for a polyphase zirconia layer were back-calculated for a range of times required to wear to a critical depth. Inputs for the model, like the friction coefficient, are taken from the tribology literature in lieu of in-reactor tribological data. Concepts of classical fretting were used as a basis, but are modified to enable the model to accommodate the complexities of the PWR environment. Factors like grid spring relaxation, pre-oxidation of the cladding, multiple oxide phases, gap formation, impact, and hydrogen embrittlement are part of the problem definition but uncertainties in their relative roles limits the ability to validate the model. Sample calculations of wear depth versus time in the cladding illustrate how GTRF wear might occur in a discontinuous fashion during months-long reactor operating cycles. A means to account for grid/rod gaps

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To reconstruct a BWR type reactor into a high conversion reactor with no substantial changes for the reactor inner structure such as control rod structure. Constitution: The horizontal cross sectional shape of a channel box is reformed into a square configuration and the arrangement of fuel rods is formed as a trigonal lattice-like configuration. As a method of improving the conversion ratio, there is considered to use a dense lattice by narrowing the distance between fuel rods and trigonal lattice arrangement for fuel rod is advantageous therefor. A square shape cross sectional configuration having equal length both in the lateral and longitudinal directions is suitable for the channel box as a guide upon movement of the control rod. Fuel rods can be arranged with no loss by the trigonal lattice configuration, by which it is possible to improve the neutron moderation, increase the reactor core reactivity and conduct effective fuel combustion. In this way, it is possible to attain the object by inserting the follower portion of the control rod at the earier half and extracting the same at the latter half during the operation period in the reactor core comprising fuel assemblies suitable to a high conversion BWR type reactor having average conversion ratio of about 0.8. (Kamimura, M.)

  13. Analysis of the performance of fuel cells BWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles BWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Vargas, S.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: govaj666@hotmail.com

    2008-07-01

    The efficient use of the fuel is one of the objectives in the assemblies design of type BWR. The present tendency in the assemblies design of type BWR is through a radial distribution of enrichments. The present work has like object showing the because of this decision, for what a comparison of the neutronic performance of two fuel cells with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The cells were analyzed with the CASMO-4 code and the obtained results of the behavior of the neutron flow and the power sustain the because of the radial distribution of enrichments. (Author)

  14. Encapsulation of failed fuel rods - a Siemens contribution to the customer's back end services

    International Nuclear Information System (INIS)

    In order to assist the customer's need to avoid water contamination of the spent fuel pool in case of defective irradiated fuel assemblies, and to assure the defect-free intermediate storage of spent fuel assemblies, Siemens has developed techniques to encapsulate defective fuel rods in special gas-tight capsules. The design features, handling, and performance of the Siemens capsules for defective fuel rods are outlined. The experience of the first encapsulation campaign of 27 defective fuel rods of different suppliers in the Finnish Boiling Water Reactors Olkiluoto 1 and 2 is also described. And an outlook of further Siemens activities in the field of gas-tight and particle-tight capsules and canisters for defective fuel rods in Boiling Water Reactors or Pressurized Water Reactors is given. (authors)

  15. Nuclear fuel rod straightness measuring system and method

    International Nuclear Information System (INIS)

    A method is described for measuring the straightness of a rod, comprising the following steps: (a) supporting the rod so that if the rod were straight, the rod would remain straight without transverse translational movement while supported and if rotated about its longitudinal axis, and so that if the rod were cambered, the rod would remain so cambered while supported and if rotated; (b) rotating the supported the rod so that if the rod were straight, the rod would be rotated about its longitudinal axis; (c) measuring the distances during the rotation between the supported and rotating the rod and rigidly-mounted, spaced-apart range finders, the range finders disposed apart from and directed towards the supported and rotating the rod and disposed so that if the rod were straight, the range finders would each be directed transverse to the longitudinal axis; and (d) calculating for each of the range finders the difference between the maximum and minimum of the distance measurements, the differences indicating the degree of straightness of the rod

  16. A new SCWR fuel assembly with two-row fuel rods between the hexagonal moderator channels

    International Nuclear Information System (INIS)

    Highlights: • We propose a SCWR fuel assembly with two-row fuel rods between the hexagonal moderator channels. • The new concept can resolve the contradiction between uniform and sufficient moderation. • Structural size and thermal–hydraulic performance are taken account of in the fuel assembly. • Larger infinite multiplication factor and smaller local power peaking factor could be obtained. • Two two-row hexagonal fuel assembly concepts are proposed for the engineering application. - Abstract: A new hexagonal fuel assembly (FA) design which has two rows of fuel rods between the hexagonal moderator channels is proposed for the thermal supercritical water cooled reactor (SCWR). The new concept is well considered for the performance of uniform moderation and sufficient moderation, and with respect to structural size and thermal–hydraulic performance. The neutron physical performance of the two-row hexagonal FA with acceptable configuration is discussed. The results show clearly that a better balance between uniform moderation and sufficient moderation can be obtained in the two-row hexagonal fuel assembly

  17. Operation method for BWR type reactor

    International Nuclear Information System (INIS)

    In a BWR type reactor, the number of fuels at low enrichment, among initially loaded fuels, is increased greater than that of fuels to be exchanged, and the number of fuels at low enrichment remained in a reactor core after fuel exchange is decreased to smaller than that of entire control rods. Further, the fuels at low enrichment are disposed to the inner side except for the outermost circumference in the reactor core after fuel exchange. Since fuels of high reactivity are disposed at the outermost circumference in a second cycle, leakage of neutrons is increased and effective breeding factor is decreased. However, since the number of brought over fuels at low enrichment is decreased and the number of fuels at high enrichment is increased, effective average reactor core enrichment degree is increased, to compensate the lowering thereof due to the increase of neutron leakage. Since dispersion effect for the distribution of the enrichment degree can be utilized as much as possible by greatly reducing the number and the enrichment degree of fuels at low enrichment for initially loaded fuels, irrespective of the average enrichment degree and the fueling pattern in a first cycle, a burnup degree upon take-out of initially loaded fuels at ow enrichment degree can be increased to maximum. (N.H.)

  18. End plug for fuel rod and welding method therefor

    International Nuclear Information System (INIS)

    An end plug of a fuel rod comprises a pressure-insertion portion having a diameter somewhat greater than the inner diameter of a fuel cladding tube and a welding portion having a diameter substantially the same as the outer diameter of the cladding tube. A V-shaped recess having an outer diameter smaller than the greatest outer diameter of the pressure-insertion portion is formed over the entire circumferential surface of the outer circumference of the connection portion of the pressure-insertion portion and the welding portion. The pressure-insertion portion of the end plug is inserted to the end of the cladding tube till the end of the cladding tube abuts against the inclined surface of the V-shaped recess. The abutting surfaces of the end plug and the cladding tube are subjected to resistance welding in this state. The inner portion bulged from the inclined surface of the V-shaped recess is filled in the recess in a molten state. Lowering of temperature of the cladding tube in the vicinity of the welded portion is decreased by γ heat during reactor operation. Accordingly, lowering of ductility of the cladding tube and degradation of material of the welded region due to segregation of hydrogen in the cladding tube can be suppressed. (I.N.)

  19. The effects of fission gas release on PWR fuel rod design and performance

    International Nuclear Information System (INIS)

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50 000 to 60 000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  20. The effects of fission gas release on PWR fuel rod design and performance

    International Nuclear Information System (INIS)

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50,000 to 60,000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  1. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    Directory of Open Access Journals (Sweden)

    Alroumi Fawaz

    2016-01-01

    Full Text Available Control rod reactivity (worths for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is widely benchmarked for various reactor types and complexities in their geometric arrangements of the assemblies and reactor core material distributions. Thus, it is used as a base methodology to evaluate neutronics variables of the research reactor at the University of Utah. With its much shorter computation time than MCNP6, AGENT provides agreement with the MCNP6 within a 0.5 % difference for the eigenvalue and a maximum difference of 10% in the power peaking factor values. Differential and integral control rod worths obtained by AGENT show well agreement with MCNP6 and the theoretical model. However, regulating the control rod worth is somewhat overestimated by both MCNP6 and AGENT models when compared to the experimental/theoretical values. In comparison to MCNP6, the total control rod worths and shutdown margin obtained with AGENT show better agreement to the experimental values.

  2. Uncertainty analysis of spent nuclear fuel isotopics and rod internal pressure

    Science.gov (United States)

    Bratton, Ryan N.

    The bias and uncertainty in fuel isotopic calculations for a well-defined radio- chemical assay benchmark are investigated with Sampler, the new sampling-based uncertainty quantification tool in the SCALE code system. Isotopic predictions are compared to measurements of fuel rod MKP109 of assembly D047 from the Calvert Cliffs Unit 1 core at three axial locations, representing a range of discharged fuel burnups. A methodology is developed which quantifies the significance of input parameter uncertainties and modeling decisions on isotopic prediction by compar- ing to isotopic measurement uncertainties. The SCALE Sampler model of the D047 assembly incorporates input parameter uncertainties for key input data such as multigroup cross sections, decay constants, fission product yields, the cladding thickness, and the power history for fuel rod MKP109. The effects of each set of input parameter uncertainty on the uncertainty of isotopic predictions have been quantified. In this work, isotopic prediction biases are identified and an investiga- tion into their sources is proposed; namely, biases have been identified for certain plutonium, europium, and gadolinium isotopes for all three axial locations. More- over, isotopic prediction uncertainty resulting from only nuclear data is found to be greatest for Eu-154, Gd-154, and Gd-160. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle as- sembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each considered WBN1 fuel rod. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burn- able absorber (IFBA) layers is

  3. Analysis of the Behavior of CAREM-25 Fuel Rods Using Computer Code BACO

    International Nuclear Information System (INIS)

    The thermo-mechanical behavior of a fuel rod subjected to irradiation is a complex process, on which a great quantity of interrelated physical-chemical phenomena are coupled.The code BACO simulates the thermo-mechanical behavior and the evolution of fission gases of a cylindrical rod in operation.The power history of fuel rods, arising from neutronic calculations, is the program input.The code calculates, among others, the temperature distribution and the principal stresses in the pellet and cladding, changes in the porosity and restructuring of pellet, the fission gases release, evolution of the internal gas pressure.In this work some of design limits of CAREM-25's fuel rods are analyzed by means of the computer code BACO.The main variables directly related with the integrity of the fuel rod are: Maximum temperature of pellet; Cladding hoop stresses; Gases pressure in the fuel rod; Cladding axial and radial strains, etc.The analysis of results indicates that, under normal operation conditions, the maximum fuel pellet temperature, cladding stresses, pressure of gases at end of life, etc, are below the design limits considered for the fuel rod of CAREM-25 reactor

  4. The formation process of the pellet-cladding bonding layer in high burnup BWR fuels

    International Nuclear Information System (INIS)

    The bonding formation process was studied by EPMA analysis, XRD measurements, and SEM/TEM observations for the oxide layer on a cladding inner surface and the pellet-cladding bonding layer in irradiated fuel rods. Specimens were prepared from fuels which had been irradiated to the pellet average burnups of 15, 27, 42 and 49 GWd/t in BWRs. In the lower burnup specimens of 15 and 27 GWd/t, no bonding layer was found, while the higher burnup specimens of 42 and 49 GWd/t had a typical bonding layer about 10 to 20 μm thick. A bonding layer which consisted of two regions was found in the latter fuels. One region of the inner surface of the Zr liner cladding was made up mainly of ZrO2 with a small amount of dissolved UO2. The structure of this ZrO2 consisted of cubic polycrystals a few nanometers in size, while no monoclinic crystals were found. The other region, near the pellet surface, had both a cubic solid solution of (U,Zr)O2 and amorphous phase in which the concentrations of UO2 and ZrO2 changed continuously. Even in the lower burnup specimens having no bonding layer, cubic ZrO2 phase was identified in the cladding inner oxide layer. The XRD measurements were consistent with the TEM results of the absence of the monoclinic ZrO2 phase. Phase transformation and amorphization were attributed to fission damage, since such phenomena have never been observed in the cladding outer surface. Phase transformation from monoclinic to cubic ZrO2 and amorphization by irradiation damage of fission products were discussed in connection with the formation mechanism and conditions of the bonding layer. (author)

  5. Design and evaluation of an on-line fuel rod assay device for an HTGR fuel refabrication plant

    International Nuclear Information System (INIS)

    Refabricated HTGR fuel rods will contain from approx. 0.15 to 0.5 g 233U and/or 235U. The fuel rods are approx. 16 mm in diameter and 62 mm long. A typical commercial fuel refabrication facility will have six fuel rod production lines, each producing approximately one fuel rod every 4 seconds at design capacity. One on-line assay device will be present for each two production lines. The relative standard deviation in an individual fuel rod fissile material measurement must be less than 3% to satisfy process and quality control requirements. Systematic errors must be kept less than approx. 0.3% for fissile material measured in fuel rods produced over two months to satisfy material accountability requirements. Several nondestructive assay (NDA) methods were investigated. Because the gamma-ray activity of the refabricated fuel is relatively high due to the presence of 232U in the fuel and because the gamma-ray activity is not directly related to total or fissile uranium content, NDA methods employing gamma-ray detection did not appear practicable. A method using thermal neutron irradiation and fast-fission neutron detection was selected. An experimental assay device was fabricated based on this NDA method. Experiments were performed to determine the precision and accuracy of the measurements and to investigate potential interferences and systematic errors. Operating procedures were evaluated, and analysis procedures were identified

  6. Estimation of water-water energy reactor fuel rod failure in design basis accidents

    International Nuclear Information System (INIS)

    The definition of failure fuel rod amount under water-water energy reactor (WWER) design basis accidents (DBA) conditions is an urgent task of modern design substantiations of WWER type fuel cycles, it is necessary for an adequate estimation of possible radiological consequences of DBA. The various aspects of a problem devoted to definition of failure fuel rod quantity under WWER DBA are considered: procedural, experimental, settlement-analytical. To analyze fuel rod behavior and to forecast by settlement cladding failure under DBA conditions (loss of coolant accident (LOCA) and reactivity initiated accident (RIA)) the RAPTA-5 code is used. For support and development of the RAPTA-5 program the experimental researches results of WWER fuel rod behavior under conditions, characterized for LOCA and RIA are used. The growing requirements of modern design substantiations cause necessity of thermal-mechanical and corrosion fuel rod models specifications, decrease of models conservatism, expansion of applicability ranges concerning fuel burnup, fuel and cladding materials, conditions of fuel rod loading. In pile and out of pile experiments, which were used for models development and verification of the RAPTA-5 code, are submitted. For account of cladding plastic deformation the multi-parametric function of a cladding material flow stress depended upon strain and strain rate, temperature and heating rate, fast neutrons fluence, oxygen concentration is used. To determine realistic estimations of cladding hoop strain at failure moment the non-axis-symmetrical deformation model of fuel rod cladding is proposed. The verification of the given model is carried out: by test results of WWER-1000 type 37-fuel rods assembly with E110 cladding on the electro-heating PARAMETER - M facility, the temperature mode of fuel rod cladding under second stage of LOCA conditions was simulated in this experiment; by test results of BT-2 experiment, performed on the MIR research reactor, where

  7. Numerical analysis on the dynamic response of nuclear fuel rods under side drop loading

    International Nuclear Information System (INIS)

    Spent nuclear fuel assemblies are stored in the cask with the necessary protective function during the spent fuel transportation or storage. The structural integrity of the fuel rods for the impact load caused by cask mishandling drop accident has to be evaluated to realize more economical and safe operation. However, the experiments for the spent fuel rods accompany many of difficulties and the available experimental data is limited. Hence, the numerical simulation is useful method to predict and evaluate the dynamic response of the fuel rods. This paper deals with the dynamic response of the spent fuel rods under side drop condition. One spacer span of the fuel rod of the undermost layer during side drop is analyzed by finite element method (LS-DYNA). The fuel rod was held between the upper and lower spacers. In the numerical model, the fixed rigid base and the drop weight were connected to the lower and upper spacers, respectively. The numerical analyses were carried out by varying the impact speed of the drop weight and the spacer plate size. Further, the cases of entire drop, in which both the drop weight and fuel rod have the same impact speed, and weight drop, in which only the weight collides with the spent fuel rod placed on the rigid wall were also analyzed. It is found that the fuel pellet and cladding tube reached their yield stresses near the impact point even for the lowest impact speed (2.4 m/s) regardless of the spacer size and drop conditions. The plastic region of the pellets is limited within one pellet and the corner part of the neighboring pellet even for the highest impact speed (13.3 m/s). The deflection of the fuel rod increased from impact point to the end of the fuel rod in the entire drop condition, but it was limited near the impact point in the weight drop condition. Therefore, the weight drop condition may underestimate the deformation or damage of the spent fuel rod compared to the entire drop condition. (author)

  8. Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

    International Nuclear Information System (INIS)

    This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

  9. Adaptive fuzzy system for fuel rod cladding failure in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Guimaraes, Antonio C.F. [Instituto de Engenharia Nuclear - Divisao de Reatores/CNEN, Ilha do Fundao s/n, 21945-970, P.O. Box 68550, Rio de Janeiro (Brazil)]. E-mail: tony@ien.gov.br; Lapa, Celso M.F. [Instituto de Engenharia Nuclear - Divisao de Reatores/CNEN, Ilha do Fundao s/n, 21945-970, P.O. Box 68550, Rio de Janeiro (Brazil)]. E-mail: lapa@ien.gov.br

    2007-03-15

    A new approach to the study of ballooning that causes cladding failure in fuel rods using an adaptive neural fuzzy inference system (ANFIS) is presented in this paper. By mapping input/output patterns describing cladding failure phenomena through average inner cladding temperature and fuel rod gas pressure, ANFIS shows a great potential to modeling this problem in alternative to the traditional approach. A typical pressurized water reactor fuel rod data was used to this application. The results confirm the potential of ANFIS comparatively to experimental calculations.

  10. Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

    Energy Technology Data Exchange (ETDEWEB)

    Rowsell, David Leon [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-06-01

    This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

  11. Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods (LWBR Development Program)

    International Nuclear Information System (INIS)

    Twenty-one LWBR irradiation test rods containing ThO2-UO2 fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxide thickness of non-defected rods, gave results which were in reasonable agreement with the outer surface oxide thicknesses of defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate and to time, the calculated values agreed well with measured inner oxide corrosion film values. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods. 16 refs., 6 figs., 8 tabs

  12. Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J.C.

    1985-08-01

    Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxide thickness of non-defected rods, gave results which were in reasonable agreement with the outer surface oxide thicknesses of defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate and to time, the calculated values agreed well with measured inner oxide corrosion film values. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods. 16 refs., 6 figs., 8 tabs.

  13. Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

    International Nuclear Information System (INIS)

    This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

  14. Rod consolidation of RG and E's [Rochester Gas and Electric Corporation] spent PWR [pressurized water reactor] fuel

    International Nuclear Information System (INIS)

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister

  15. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J C

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated.

  16. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    International Nuclear Information System (INIS)

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

  17. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  18. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

  19. Shielding analysis of the transfer container and the borehole lock for spent fuel rod final storage in BSK 3; Abschirmanalysen von Transferbehaelter und Bohrlochschleuse zur Endlagerung abgebrannter Brennstaebe in BSK 3

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Fopp, S.; Graf, R. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Filbert, W. [DBETEC DBE Technology GmbH, Peine (Germany)

    2008-07-01

    In the frame of ESDRED (engineering studies and demonstration of repository designs) DBE Technology and GNS enhance the concept of final storage of spent fuel rods in un-shielded canisters (BSK 3) in boreholes of salt formations. The transfer container (TSB) is envisaged as radiological shielding for the transport of BSK 3 within the facility. The contribution covers the radiological shielding calculations for the transfer container and the borehole lock for normal operation and accident condition. The inventory of BSK 3 includes fuel rods from 3 PWR or 9 BWR reactor cores, the calculations are performed using the program MCNP. The TFB is supposed to shield the BSK 3 during transport from aboveground to the borehole in the underground facility. The distribution of dose outputs along the container and at the cover ends are demonstrated to within the limits. 90% of the dose output is neutron radiation.

  20. Fuel rod behavior under normal operating conditions in Super Fast Reactor with high power density

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Haitao, E-mail: haitaoju@gmail.com [Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan 610041 (China); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, The University of Tokyo, Hongo, Bunkyo, Tokyo 113-8656 (Japan); Oka, Yoshiaki [Joint Department of Nuclear Energy, Waseda University, Totsukamachi, Shinjuku, Tokyo 169-8050 (Japan)

    2015-08-15

    Highlights: • The improved core of Super Fast Reactor with high power density is analyzed. • We analyzed four types of the limiting fuel rods. • The influence of Pu enrichment and compressive stress to yield strength ratio are analyzed. • The improved fuel rod design of the new core is suggested. - Abstract: A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8 W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8 W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumferential direction should

  1. Investigating mechanical behavior and radiation resistant of fuel rods clad in nuclear power plant

    International Nuclear Information System (INIS)

    interstitials in metal lattice under irradiation causes increased strength and hardness but decreases ductility in metals.The increase in strength and hardness depends on obstacles that prevent the motion of dislocations. The clustering of point defects are responsible for these changes. Irradiation also induces instabilities in phases due to enhancement of diffusion, solute segregation, precipitate formation, order- disorder transformation and resolution of small precipitates. From the microscopic point of view accumulation of vacancies accompanied by formation of He and H2 gases under irradiation cause an increase in volume which results in swelling and eventually ends up with embrittlement of metals. This subject was described in chapter three Zirconium and its alloys are the best structural materials for fuel cladding of BWR and PWR reactors core. The working condition in the core of nuclear reactor are very serve, respect temperature and radiation dose. It should be realized that, if fuel cladding receive damage and get cracked, the first cooling cycle and the maine equipment will be contaminated with active materials which cause additional environmental problems. Furthermore, replacement of fuel rods are very costly. Therefore, for increasing life time of fuel cladding and minimizing damage, the effect of radiation and heat on Zirconium and its alloys must be investigated. This subject was described in chapter four.The mechanical behavior and radiation resistant of fuel cladding in PWR reactor (specifically WWER ) have been investigated which is described in chapter five. Result, discussion and final conclusion are summarized in last chapter and also several points for improvement have been offered

  2. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  3. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    International Nuclear Information System (INIS)

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL

  4. Development of oxide layer thickness measuring device for irradiated nuclear fuel rods in hot cell

    International Nuclear Information System (INIS)

    It has been known that water side corrosion of fuel rods in nuclear reactor is accompanied with the loss of metallic wall thickness and pickup of hydrogen. This corrosion is one of the important limiting factors in the operating life of fuel rods. In connection with the fuel cladding corrosion, a device to measure the water side oxide layer thickness by means of the eddy current method without destructing the fuel rod was developed by KAERI. The device was installed on the multi function testing bench in the nondestructive test hot cell and its calibration was carried out successfully for the standard rod attached with plastic thin films whose thicknesses are predetermined. It shows good precision within 10% error

  5. Seal-welding detection device for fuel rod and detection method therefor

    International Nuclear Information System (INIS)

    The present invention provides a method of and a device for detecting presence or absence of abnormality of welded portions (nugget portions) for sealing an end plug sealing hole of a fuel rod. Namely, the end face of the fuel rod is photographed, and the nugget region is detected based on the photographed images by a nugget boundary recognition means. The region to be sealed is determined as a control-range mark on the image of the end face of the fuel rod. Whether the control-range mark is included in the nugget portion or not is compared and evaluated as a comparing and judging means. Then, the presence or absence of the abnormality for the seal in the nugget portion of the fuel rod end plug can be visually monitored automatically and continuously. In addition, since the remote-detection can be conducted and in non-contact manner, operator's exposure can be eliminated. (I.S.)

  6. Safety analysis report supplement for consolidated fuel rod storage at an independent spent fuel storage installation. Volume IV

    International Nuclear Information System (INIS)

    The purpose of this report is to supplement a safety analysis report for an independent spent fuel storage installation to allow the incorporation of consolidated nuclear fuel into an otherwise licensed facility. This report presents only information regarding storage of consolidated pressurized water reactor fuel. The document contains supplemental criticality, radiological, structural, and thermal information which must be added to an existing safety analysis report to permit upgrading the method of fuel storage to consolidated fuel rod storage

  7. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  8. Water-side oxide layer thickness measurement of the irradiated PWR fuel rod by NDT method

    International Nuclear Information System (INIS)

    It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the loss of metallic wall thickness and pickup of hydrogen. This corrosion is one of the important limiting factors in the operating life of fuel rods. In connection with the fuel cladding corrosion, a device to measure the water-side oxide layer thickness by means of the eddy-current method without destructing the fuel rod was developed by KAERI. The device was installed on the multi-function testing bench in the nondestructive test hot-cell and its calibration was carried out successfully for the standard rod attached with plastic thin films whose thicknesses are predetermined. It shows good precision within about 10% error. And a PWR fuel rod, one of the J-44 assembly discharged from Kori nuclear power plant Unit-2, has been selected for oxide layer thickness measurements. With the result of data analysis, it appeared that the oxide layer thicknesses of Zircaloy cladding vary with the length of the fuel rod, and their thicknesses were compared with those of the destructive test results to confirm the real thicknesses

  9. Non-destructive testing dummy nuclear fuel rods by neutron radiography

    International Nuclear Information System (INIS)

    Background: The nuclear fuel rod is a key component of nuclear plants and reactors. It works in the extreme conditions, so it is easy to be broken. In order to be safe in operation, lots of testings have to be carried out from fabricating to operating of the fuel rod. Purpose: As a unique non-destructive testing technique, neutron radiography can be used to measure the nuclear fuel rods with radioactivity by an indirect neutron radiography method. Study the indirect neutron radiography method is the primary step of testing. Methods: Non-destructive testing experiments were carried out at China Advanced Research Reactor (CARR) by indirect neutron radiography method with dummy nuclear fuel rods as the samples. The 0.1 mm-thick Dy foil was used as the neutron converter. Results: The neutron images of dummy nuclear fuel rods were obtained. The resolution of testing was analyzed with the images. Through imaging analysis methods, the structure defections, the hydrogen accumulation in the cladding and the U-235 enrichment of pellet were studied and analyzed. Conclusions: The indirect neutron radiography method and the neutron image analysis method were studied. The work described in this paper provides a primary guideline for investigating actual irradiated fuel rods by the neutron radiography at CARR in the future. (authors)

  10. Method and apparatus for reading lased bar codes on shiny-finished fuel rod cladding tubes

    International Nuclear Information System (INIS)

    This patent describes, in a nuclear fuel rod identification system, a method of reading a bar code etched directly on a surface of a nuclear fuel rod. It comprises: defining a pair of light diffuser surfaces adjacent one another but in oppositely inclined relation to a beam of light emitted from a light reader; positioning a fuel rod, having a cylindrical surface portion with a bar code etched directly thereon, relative to the light diffuser surfaces such that the surfaces are disposed adjacent to and in oppositely inclined relation along opposite sides of the fuel rod surface portion and the fuel rod surface portion is aligned with the beam of light emitted from the light reader; directing the beam of light on the bar code on fuel rod cylindrical surface portion such that the light is reflected therefrom onto one of the light diffuser surfaces; and receiving and reading the reflected light from the bar code via the one of the light diffuser surfaces to the light reader

  11. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    International Nuclear Information System (INIS)

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 25800F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs

  12. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Richardson, K.D.

    1987-10-01

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.

  13. BWR spent fuel transport and storage system for KKL: TN trademark 52L, TN trademark 97L, TN trademark 24 BHL

    Energy Technology Data Exchange (ETDEWEB)

    Sicard, D.; Verdier, A. [COGEMA Logistics (AREVA Group) (France); Monsigny, P.A. [NOK/KKL (Switzerland)

    2004-07-01

    The LEIBSTADT (KKL) nuclear power plant in Switzerland has opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN trademark a52L and TN trademark 97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TNae24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN trademark 52L and TN trademark 97L casks by the KKL and ZWILAG operators.

  14. BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR; BUTREN-RC un sistema hibrido para la optimizacion de recargas de combustible nuclear en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Castillo M, J.A. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico); Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

  15. NSR-77: a computer code for transient analysis of a light water reactor fuel rod

    International Nuclear Information System (INIS)

    This report describes computer code NSR-77 written in FORTRAN IV for FACOM-M 200 computer in detail. It has been developed for transient response analysis of a light water reactor fuel rod during an accident such as a reactivityy initiated accident, a loss-of-coolant accident or a power-cooling-mismatch accident. The code consists of subcodes which calculate heat conduction in a fuel rod, gas gap conductance between fuel and cladding, heat transfer from cladding to coolant, fluid hydrodynamics, elastic-plastic fuel and cladding deformation, and material properties, and so on. (author)

  16. Development and control of the process for the manufacture of zircaloy-4 tubing for LWBR fuel rods

    International Nuclear Information System (INIS)

    The technical requirements for the Light Water Breeder Reactor (LWBR) fuel elements (fuel rods) imposed certain unique requirements for the low hafnium Zircaloy-4 tubing used as fuel rod cladding. This report describes, in detail, the tube manufacturing process, the product and process controls used, the inspections and tests performed, and the efforts involved in refining a commercial tube reducing process to produce tubes that would satisfy the requirements for LWBR fuel rod cladding

  17. Results from studies of surface deposits on the claddings of fuel rods used in RBMK-1000 reactors

    Science.gov (United States)

    Smirnova, I. M.; Markov, D. V.

    2010-07-01

    The results of studies on analyzing the element composition of deposits on the cladding surfaces of fuel rods used in a fuel assembly at the Leningrad nuclear power station are presented. The distribution of elements in deposits over the fuel rod height is analyzed, and the zones of their concentration are revealed. It is shown that deposits of copper penetrating into cracks in the surface layer of zirconium oxide introduce an essential contribution in the development of nodular corrosion of fuel rod claddings.

  18. Qualification Test of End Closure Welding for SFR Metallic Fuel Rod

    International Nuclear Information System (INIS)

    The six systems selected for Gen-IV promising systems and expected to be deployed in 2020. In Korea, the R and D on a sodium-cooled fast reactor (SFR) has been begun since 1997, as one of the national long-term nuclear R and D programs. As a fuel for SFR, metallic fuel, U-Zr alloy fuel, was selected and is being developed. And also the international collaborative research is under way on U-TRU-Zr fuel developments with the closed fuel cycle of full actinide recycling, within Advanced Fuel Project for international generation IV (Gen-IV) SFR. For the fabrication of SFR metallic fuel rods, the end plug welding is a crucial process. The sealing of end plug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the qualification test of the end closure welding for SFR metallic fuel rods was carried out based on the developed welding technique, welding equipment, welding conditions and parameters. Through the qualification test, the weld quality of end closure welding of SFR metallic fuel rod was evaluated according to the qualification test plan and satisfied the requirements on the weld. Consequently the qualified welding process is ready to produce SFR metallic fuel rods

  19. The development of 3-D tomography method based on gamma-scanning of irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Dobrin, R. [RENEL, Nuclear Power Group, Inst. for Nuclear Research (INR), Post-Irradiation Examination Lab., Pitesti (Romania); Craciunescu, T. [National Inst. of Nuclear Physics and Engineering, Horia Hulubei (IFIN-HH) Lab., Bucharest, Magurele (Romania)

    1999-07-01

    The tomographic method, consisting in the reconstruction of the images from their projections, is a relatively new nondestructive method, used in the post-irradiation examination of the nuclear fuel. The purpose of the method is to determine the distribution of gamma radioactive fission products in a cross-section of an irradiated nuclear fuel rods. More than 40 fuel rods were investigated at the INR hot cell facility, using this nondestructive technique. The method is used in conjunction with the gamma scanning method and the equipment used for tomographic investigation of the irradiated nuclear fuel is the same to that usually used for gamma scanning investigation. The paper presents the principles of the Gamma Emission Computed Tomography (GECT), as well as the results and performance of this method, when applied to the investigation of some types of fuel rods, irradiated in the TRIGA 14 MW{sub th} materials testing reactor. (author)

  20. Ultrasonic testing of nuclear fuel rod welds and clad (LWBR Development Program)

    International Nuclear Information System (INIS)

    Ultrasonic techniques were developed utilizing commercially available equipment as a part of the work required in the LWBR Core Manufacturing program for assurance of fuel rod weld integrity and for measurement of fuel rod clad thickness and clad thickness eccentricity. The need for the highest possible resolution and the undesirability of transducer to rod contact dictated the use of a water immersion technique with pulse-echo instrumentation for both weld and clad thickness ultrasonic tests. For the weld test, both longitudinal wave and shear wave inspections were employed with the transducers shuttled together back and forth longitudinally across the weld zone as the fuel rod was rotated. For the clad thickness test, only the longitudinal wave inspection was used with a helical scan pattern along the full length of the clad

  1. Optimization of analysis best-estimate of a fuel element BWR with Code STAR-CCM+; Optimizacion del analisis best-estimate de un elemento combustible BWR con el codigo STAR-CCM+

    Energy Technology Data Exchange (ETDEWEB)

    Morgado Canada, E.; Concejal Barmejo, A.; Jimenez Varas, G.; Solar Martinez, A.

    2014-07-01

    The objective of the project is the evaluation of the code STAR-CCM +, as well as the establishment of guidelines and standardized procedures for the discretization of the area of study and the selection of physical models suitable for the simulation of BWR fuel. For this purpose several of BFBT experiments have simulated [1] provide a data base for the development of experiments for measuring distribution of fractions of holes to changes in power in order to find the most appropriate models for the simulation of the problem. (Author)

  2. A two-dimensional finite element method for analysis of solid body contact problems in fuel rod mechanics

    International Nuclear Information System (INIS)

    Two computer codes for the analysis of fuel rod behavior have been developed. Fuel rod mechanics is treated by a two-dimensional, axisymmetric finite element method. The program KONTAKT is used for detailed examinations on fuel rod sections, whereas the second program METHOD2D allows instationary calculations of whole fuel rods. The mechanical contact of fuel and cladding during heating of the fuel rod is very important for it's integrity. Both computer codes use a Newton-Raphson iteration for the solution of the nonlinear solid body contact problem. A constitutive equation is applied for the dependency of contact pressure on normal approach of the surfaces which are assumed to be rough. If friction is present on the contacting surfaces, Coulomb's friction law is used. Code validation is done by comparison with known analytical solutions for special problems. Results of the contact algorithm for an elastic ball pressing against a rigid surface are confronted with Hertzian theory. Influences of fuel-pellet geometry as well as influences of discretisation of displacements and stresses of a single fuel pellet are studied. Contact of fuel and cladding is calculated for a fuel rod section with two fuel pellets. The influence of friction forces between fuel and cladding on their axial expansion is demonstrated. By calculation of deformations and temperatures during an instationary fuel rod experiment of the CABRI-series the feasibility of two-dimensional finite element analysis of whole fuel rods is shown. (orig.)

  3. Thermohydraulics and fuel rod behavior during reflood/quench situations

    Energy Technology Data Exchange (ETDEWEB)

    Hering, W.; Homann, C.; Sanchez, V.; Sengpiel, W.; Struwe, D. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Reaktorsicherheit; Lamy, J.S. [EDF, Paris (France)

    2001-07-01

    At the Forschungszentrum Karlsruhe (FZKA) the Institute for Reactor Safety (IRS) supports the experimental activities focused on the safety features of existing and advanced light water reactors by plant calculations using SCDAP/RELAP5 (S/R5) mod3.2 and RELAP5 (R5) mod 3.3. These codes are continuously validated by comparison with various experiments. In the area of design basis safety R5 mod3.3.b code validation and code application for the high performance LWR (HPLWR, 5{sup th} European Framework Programme is ongoing), extending the capability of RELAP5 to simulate conditions of supercritical water. Best estimate safety analyses for severe accidents with the in-house version S/R5 mod 3.2.irs were performed successfully and documented for the European Pressurized Water Reactor. (FZK-6299, -6315, -6567, http://bibliothek.fzk.de/zb/berichte). Analytical support for bundle experiments in the QUENCH facility at FZK is ongoing with S/R5 mod3.2.irs to define test parameters and to support experimental analyses. A fine discretization with 5 cm meshes has been developed to better simulate reflood conditions. The improved shattering model works fairly well for an intact rod geometry, but is not adequate for damaged fuel rods. Furthermore, S/R5 is used for test analyses especially for QUENCH-06, basis for the OECD/NEA/CSNI International Standard Problem No. 45 ISP-45, and for QUENCH-07 as contribution to EU-program COLOSS. A code extension to simulate B4C oxidation is planned, but postponed until 2003 due to limited man power caused by the OECD/NEA ISP-45 benchmark. Besides, ICARE2V3 is used to support S/R5 analyses with respect to QUENCH and Phebus analyses. The blind phase of the OECD/NEA/CSNI International Standard Problem No. 45 ISP-45 (QUENCH-06) is finished, the preparation of the OECD Interpretation and Comparison report is under way. The current state of code to code comparison reveals strong variation within the same code family probably attributed to user effects

  4. Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

    Directory of Open Access Journals (Sweden)

    Giovedi Claudia

    2016-01-01

    Full Text Available Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348 and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding. Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding material.

  5. Opposed piston engine having fuel inlet through rod controlled piston port

    Energy Technology Data Exchange (ETDEWEB)

    Lively, E.P. Sr.

    1991-07-09

    This patent describes an internal combustion engine. It comprises at least one of each of an intake port, exhaust port and fuel inlet port; a pair of opposed pistons within a cylinder of the engine defining a combustion chamber; one of the pair of pistons opening and closing the at least one exhaust port, the one piston including the fuel inlet port therethrough; a connecting rod operatively connecting the one piston to a driven shaft, the connecting rod having an end portion which opens and closes the fuel inlet port.

  6. Code Package to Analyze Parameters of the WWER Fuel Rod. TOPRA-2 Code - Verification Data

    International Nuclear Information System (INIS)

    Presented are the data for computer codes to analyze WWER fuel rods, used in the WWER department of RRC 'Kurchatov Institute'. Presented is the description of TOPRA-2 code intended for the engineering analysis of thermophysical and strength parameters of the WWER fuel rod - temperature distributions along the fuel radius, gas pressures under the cladding, stresses in the cladding, etc. for the reactor operation in normal conditions. Presented are some results of the code verification against test problems and the data obtained in the experimental programs. Presented are comparison results of the calculations with TOPRA-2 and TRANSURANUS (V1M1J06) codes. Results obtained in the course of verification demonstrate possibility of application of the methodology and TOPRA-2 code for the engineering analysis of the WWER fuel rods

  7. Suitability of a thermal design method for FBR oxide fuel rods

    International Nuclear Information System (INIS)

    To study suitability of the thermal design method for fast breeder reactor (FBR) oxide fuel rods, the B14 irradiation test with four fuel rods was carried out in the experimental fast reactor 'JOYO'. Pellet-cladding gap width and O/M ratio of oxide fuels were specified as experimental parameters. In addition, by taking into account the actual design conditions for FBR oxide fuel, the conditions in the B14 irradiation test, i.e. linear power and cladding temperature, were planned to include the hottest design conditions. The maximum Pu content and the maximum Am content of fuel pellet were 31 wt% and 2.4 wt%, respectively. The B14 fuel rods were irradiated with the maximum linear power of ∼ 47 kW/m in the test. After irradiation, ceramography samples were taken from the axial position of each fuel rod where the fuel centerline temperature reached the maximum during irradiation. The result was that the influences of both pellet-cladding gap width and O/M ratio on the fuel restructuring were observed, but the fuel melting was not observed. In addition, thermal analysis code 'DIRAD' would be suitable to evaluate the thermal behavior of oxide fuels containing several percent Am, from the result of verification by using the result of the B14 irradiation test. Moreover, from the computation result of DIRAD, the power to melt for the B14 oxide fuels was evaluated as 55-57 kW/m. It could be mentioned that the margin of the conventional oxide fuel design would be at least 10 kW/m at the transient. Consequently, the margin to the criterion in the thermal design would be suitable and the fuel melting would be prevented under the conditions designated in the conventional FBR design.(author)

  8. Calculation of activity content and related properties in PWR and BWR fuel using ORIGEN 2

    International Nuclear Information System (INIS)

    This report lists the conditions for calculations of the core inventory for a PWR and BWR. The calculations have been performed using the computer code ORIGEN 2. The amount (grams), the total radioactivity (bequerels), the thermal power (watts), the radioactivity from theα-decay (bequerels), and the neutron emission (neutrons/sec) from the core after the last burnup have been determined. All the parameters have been calculated as a function of the burnup and the natural decay, the latter over a time period of 0-1.0E07 years. The calculations have been performed for 68 heavy nuclides, 60 daughter nuclides, to the heavy nuclides with atomic numbers under 92, 852 fission products and 7 light nucli ides. The most important results are listed. (author)

  9. MATPRO: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P.E.; Thompson, L.B. (eds.)

    1976-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Documentation and formulations that are generally semiempirical in nature are presented for uranium dioxide and mixed uranium-plutonium dioxide fuel, zircaloy cladding, gas mixture, and LWR fuel rod material properties.

  10. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  11. Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

    International Nuclear Information System (INIS)

    Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

  12. Fuel performance computer code simulation of steady-state and transient regimes of the stainless steel fuel rods

    International Nuclear Information System (INIS)

    The immediate cause of the accident at the Fukushima Daiichi nuclear plant in March 2011 was the meltdown of the reactor core. During this process, the zirconium cladding of the fuel reacts with water, producing a large amount of hydrogen. This hydrogen, combined with volatile radioactive materials leaked from the containment vessel and entered the building of the reactor, resulting in explosions. In the past, stainless steel was used as the coating in many pressurized water reactors (PWR) under irradiation and their performance was excellent, however, the stainless steel was replaced by a zirconium-based alloy as a coating material mainly due to its lower section shock-absorbing neutrons. Today, the stainless steel finish appears again as a possible solution for security issues related to the explosion and hydrogen production. The objective of this thesis is to discuss the performance under irradiation of fuel rods using stainless steel as a coating material. The results showed that stainless steel rods exhibit lower temperatures and higher fuel pellet width of the gap - coating the coated rods Zircaloy and this gap does not close during the irradiation. The thermal performance of the two fuel rods is very similar, and the penalty of increased absorption of neutrons due to the use of stainless steel can be offset by the combination of a small increase in the enrichment of U- 235 and changes in the size of the spacing between the fuel rods. (author)

  13. BWR simulation in a stationary state for the evaluation of fuel cell design; Simulacion de un reactor BWR en estado estacionario para la evaluacion del diseno de celdas de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  14. Water-side oxide layer thickness measurement of the irradiated PWR fuel rod by ECT method

    International Nuclear Information System (INIS)

    It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the metallic loss of wall thickness and hydrogen pickup in the fuel cladding tube. The fuel clad corrosion is one of the major factors to be controlled to maintain the fuel integrity during reactor operation. An oxide layer thickness measuring device equipped with ECT probe system was developed by KAERI, and whose performance test was carried out in NDT(Non-Destructive Test) hot-cell of PIE(Post Irradiation Examination) Facility. At first, the calibration/performance test was executed for the unirradiated standard specimen rod fabricated with several kinds of plastic thin films whose thickness were predetermined, and the result of which showed a good precision within 10% of discrepancy. And then, hot test was performed for the irradiated fuel rod selectively extracted from J44 fuel assembly discharged from Kori Unit-2. The data obtained with this device were compared with the metallographic results obtained from destructive examination in PIEF hot-cell on the same fuel rod to verify the validity of the measurement data. (author)

  15. CAST3M modelling of a spent fuel assembly bending during a handling accident Rod failure risk evaluation from the experimental results of spent fuel rod bending test

    International Nuclear Information System (INIS)

    The fuel handling operating rules exclude any accidental risk. However in the framework of the PRECCI R and D project, the bending of a spent fuel assembly resulting from its locking during a translation displacement is taken into account. This enabled us to develop an approach based on experiments and calculations that allows us to simulate the behaviour of an assembly under such loading. This study was carried out in CEA laboratories with the funding and the technical support of EDF. A three points bending test on a spent fuel rod segment was performed at the Laboratory for Mechanical Behaviour of Irradiated Materials (LCMI). From the experimental strength-displacement curve, a maximum failure strain, a maximum failure curvature and an equivalent constitutive equation were determined. CAST3M modelling of the fuel rod taking into account the elasto-plastic behaviour of the clad and the cracking of the UO2 fuel pellets was verified by the experimental results. Consequently, the identification of the respective contributions of the clad and of the pellets to the rod global behaviour was made possible. A two dimensional assembly with beam elements was modelled with CAST3M. The properties of the beams modelling the different parts of the assembly (top and bottom nozzle, grids) were chosen and adjusted according to their materials (zirconium alloys, steel) in order to obtain stiffness, tensile and shear behaviour, sliding and holding functions close to the experimental ones. Assembly bending calculations were performed. In order to obtain a rod integrity estimator, their maximum calculated strains and curvatures as a function of the bending angles can be compared to the failure experimental ones. (authors)

  16. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

  17. Actinides record, power calculations and activity for present isotopes in the spent fuel of a BWR; Historial de actinidos y calculos de potencia y actividad para isotopos presentes en el combustible gastado de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Lucatero, M. A., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The administration of spent fuel is one of the more important stages of the nuclear fuel cycle, and this has become a problem of supreme importance in countries that possess nuclear reactors. Due to this in this work, the study on the actinides record and present fission products to the discharge of the irradiated fuel in a light water reactor type BWR is shown, to quantify the power and activity that emit to the discharge and during the cooling time. The analysis was realized on a fuel assembly type 10 x 10 with an enrichment average of 3.69 wt % in U-235 and the assembly simulation assumes four cycles of operation of 18 months each one and presents an exposition of 47 G Wd/Tm to the discharge. The module OrigenArp of the Scale 6 code is the computation tool used for the assembly simulation and to obtain the results on the actinides record presents to the fuel discharge. The study covers the following points: a) Obtaining of the plutonium vector used in the fuel production of mixed oxides, and b) Power calculation and activity for present actinides to the discharge. The results presented in this work, correspond at the same time immediate of discharge (0 years) and to a cooling stage in the irradiated fuel pool (5 years). (Author)

  18. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  19. Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide; Analisis del impacto de modificar la composicion del combustible nuclear de un BWR con oxido de berilio

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo V, J. M.; Morales S, J. B., E-mail: euqrop@hotmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2013-10-15

    The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO{sub 2}) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO{sub 2} mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

  20. Fabrication of an instrumented fuel rod mock-up using a precise drilling machine

    International Nuclear Information System (INIS)

    When a new nuclear fuel is developed, Irradiation test needs to carried out in the research reactor to analyze the performance of the new nuclear fuel. In addition, to check the performance of the nuclear fuel during the burn up test in the test loop, it is necessary to attach sensors near the fuel rod and connect them with instrumentation cables to the measuring device located outside of the reactor pool. In particular, to check the temperature fluctuation of a nuclear fuel during the irradiation test, a thermocouple should be instrumented at the center of the nuclear fuel. Therefore, A hole needs to be made at the center of a fuel pellet to put in the thermocouple. However, because the hardness and density of a sintered UO2 pellet are very high, it is difficult to make a small fine hole on the sintered UO2 pellet with a simple drilling machine. In this study, an instrumented fuel rod mock-up was fabricated using an automated precise drilling machine. Four sintered alumina were drilled off and assembled into the zircaloy tube and a K-type thermocouple was instrumented in the fuel rod mock-up

  1. Process for reprocessing a nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    In order to separate the nuclear fuel from the can material, the can is heated evenly in the gastight closed state together with the nuclear fuel contained in it, so that the diameter of the can expands, increasing the gap between the nuclear fuel and the can without cracks occurring in the can. The expanded can is then opened at one end and finally the nuclear fuel from the opened can and is treated separately from the can. (orig./HP)

  2. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    International Nuclear Information System (INIS)

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab

  3. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

  4. RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

    International Nuclear Information System (INIS)

    The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The essential characteristics of the code are briefly outlined here. The model is particularly designed to perform a full thermal and mechanical analysis in both the azimuthal and radial directions. Thus, azimuthal temperature gradients arising from pellet eccentricity, flux tilt, arbitrary distribution of heat sources in the fuel and the cladding and azimuthal variation of coolant conditions can be treated. The code combines a transient 2-dimensional heat conduction code and a 1-dimentional mechanical model for the cladding deformation. The fuel rod is divided into a number of axial sections and a detailed thermomechanical analysis is performed within each section in radial and azimuthal directions. In the following sections, instructions are given for the definition of the data files and the semi-variable dimensions. Then follows a complete description of the input data. Finally, the restart option is described

  5. Experimental Investigation on Flow-Induced Vibration of Fuel Rods in Supercritical Water Loop

    Directory of Open Access Journals (Sweden)

    Licun Wu

    2014-01-01

    Full Text Available The supercritical water-cooled reactor (SCWR is one of the most promising Generation IV reactors. In order to make the fuel qualification test for SCWR, a research plan is proposed to test a small scale fuel assembly in a supercritical water loop. To ensure the structure safety of fuel assembly in the loop, a flow-induced vibration experiment was carried out to investigate the vibration behavior of fuel rods, especially the vibration caused by leakage flow. From the experiment result, it can be found that: the vibration of rods is mainly caused by turbulence when flow rate is low. However, the effects of leakage flow become obvious as flow rate increases, which could changes the distribution of vibrational energy in spectrum, increasing the vibrational energy in high-frequency band. That is detrimental to the structure safety of fuel rods. Therefore, it is more reasonable to improve the design by using the spacers with blind hole, which can eliminate the leakage flow, to assemble the fuel rods in supercritical water loop. On the other hand, the experimental result could provide a benchmark for the theoretical studies to validate the applicability of boundary condition set for the leakage-flow-induced vibration.

  6. In-pile-experiments on the fuel rod behavior in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    In-pile tests with unirradiated and irradiated fuel rod specimens have been performed to investigate the influence of nuclear parameters on the mechanisms of fuel rod failure under LOCA conditions. The test specimens had an active fuel length of 50 cm. Test series G (G1, G2/3) comprised ten single-rod tests with preirradiated specimens. The average burnup was 35,000 MWd/tsub(U). This report describes the results of the destructive posttest examinations including: (a) the mechanical behavior of the Zircaloy-4 cladding, (b) the oxidation and corrosion behavior of the cladding, (c) the structural condition and microhardness of the zircaloy, (d) the mechanical behavior of the fuel, and (e) fission gas release and volume changes of the fuel. The oxidation behavior, the structural condition, and the microhardness of the cladding are compared to the results of the test series C (2,500 MWd/tsub(U) burnup) and E (8,000 MWd/tsub(U)) and of the electrically heated fuel rod simulators. Moreover, a comparison is made between the results of all test series. (orig./RW)

  7. Radial distribution of UO{sub 2} and Gd{sub 2}O{sub 3} in fuel cells of a BWR Reactor; Distribucion radial de UO{sub 2} y Gd{sub 2}O{sub 3} en celdas de combustible de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.L.; Ortiz, J.J.; Perusquia del C, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Francois, J.L.; Martin del Campo M, C. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62500 (Mexico)]. e-mail: jlmt@nuclear.inin.mx

    2008-07-01

    The fuel system that is used at the moment in a power plant based on power reactors BWR, includes as much like the one of its substantial parts to the distribution of the fissile materials like a distribution of burnt poisons within each one of the cells which they constitute the fuel assemblies, used for the energy generation. Reason why at the beginning of a new operation cycle in a reactor of this type, the reactivity of the nucleus should be compensated by the exhaustion of the assemblies that it moves away of the nucleus for their final disposition. This compensation is given by means of the introduction of the recharge fuel, starting from the UO{sub 2} enriched in U{sup 2}35, and of the Gadolinium (Gd{sub 2}O{sub 3}). The distribution of these materials not only defines the requirements of energy generation, but in certain measures also the form in that the margins will behave to the limit them thermal during the operation of the reactor. These margins must be taken into account for the safe and efficient extraction of the energy of the fuel. In this work typical fuel cells appear that are obtained by means of the use of a emulation model of an ants colony. This model allows generating from a possible inventory of values of enrichment of U{sup 2}35, as well as of concentration of Gadolinium a typical fuel cell, which consists of an arrangement of lOxlO rods, of which 92 contain U{sup 2}35, some of these rods contain a concentration of Gd{sub 2}O{sub 3} and 8 of the total contain only water. The search of each cell finishes when the value of the Local Peak Power Factor (LPPF) in the cell reaches a minimal value, or when a pre established value of iterations is reached. The cell parameters are obtained from the results of the execution of the code HELIOS, which incorporates like a part integral of the search algorithm. (Author)

  8. A PCI failure in an experimental MOX fuel rod and its sensitivity analysis

    International Nuclear Information System (INIS)

    Within our interest in studying MOX fuel performance, the irradiation of the first Argentine prototypes of PHWR MOX fuels began in 1986 with six rods fabricated at the α Facility (CNEA, Argentina). These experiences were made in the HFR-Petten reactor, Holland. The goal of this experience was to study the fuel behaviour with respect to PMCI-SCC. An experiment for extended burnup was performed with the last two MOX rods. During the experiment the final test ramp was interrupted due to a failure in the rod. The post-irradiation examinations indicated that PCI-SCC was a mechanism likely to produce the failure. At the Argentine Atomic Energy Commission (CNEA) the BACO code was developed for the simulation of a fuel rod thermo-mechanical behaviour under stationary and transient power situations. BACO includes a probability analysis within its structure. In BACO the criterion for safe operation of the fuel is based on the maximum hoop stress being below a critical value at the cladding inner surface; this is related to susceptibility to stress corrosion cracking (SCC). The parameters of the MOX irradiation, the preparation of the experiments and post-irradiation analysis were sustained by the BACO code predictions. We present in this paper an overview of the different experiences performed with the MOX fuel rods and the main findings of the post-irradiation examinations. A BACO code description, a wide set of examples which sustain the BACO code validation, and a special calculation for BU15 experiment attained using the BACO code, including a probabilistic analysis of the influence of rod parameters on performance, are included. (author)

  9. Shielding and Containment Evaluations of the NAC-LWT Cask with Tritium Burnable Poison Rods

    International Nuclear Information System (INIS)

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the U.S. Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for a shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload

  10. Analysis of heat transfer from fuel rods with externally attached thermocouples

    International Nuclear Information System (INIS)

    This paper describes the development of 2 and 3 dimensional finite element heat conduction models to simulate the behaviour of the external thermocouples attached to the LOFT fuel rods during the blowdown phase of a large break loss-of-coolant accident. To establish the model and determine the thermal coupling between the thermocouple and the fuel rod extensive use was made of two series of experiments performed at INEL in the LOFT Test Support Facility (LTSF). These experiments were high pressure reflood experiments with fluid conditions 'typical' of those seen during the bottom-up flow period of the LOFT experiments. (author)

  11. Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

    International Nuclear Information System (INIS)

    Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

  12. Comparative calculations of the WWER fuel rod thermophysical characteristics employing the TOPRA-s and the TRANSURANUS computer codes

    International Nuclear Information System (INIS)

    A short description of the TOPRA-s computer code is presented. The code is developed to calculate the thermophysical cross-section characteristics of the WWER fuel rods: fuel temperature distributions and fuel-to-cladding gap conductance. The TOPRA-s input does not require the fuel rod irradiation pre-history (time dependent distributions of linear power, fast neutron flux and coolant temperature along the rod). The required input consists of the considered cross-section data (coolant temperature, burnup, linear power) and the overall fuel rod data (burnup and linear power). TOPRA-s is included into the KASKAD code package. Some results of the TOPRA-s code validation using the SOFIT-1 and IFA-503.1 experimental data, are shown. A short description of the TRANSURANUS code for thermal and mechanical predictions of the LWR fuel rod behavior at various irradiation conditions and its version for WWER reactors, are presented. (Authors)

  13. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  14. Development of a Probabilistic Design Method for Fast Reactor Fuel Rods

    International Nuclear Information System (INIS)

    The increase of linear power and burn-up during reactor operation is considered as one measure to ensure the utility of fast reactors in the future; for this the application of annular oxide fuel is under consideration because of their availability for both high power and high burn-up. The CEPTAR code was developed as a design code for annular oxide fuels and results of verifications using the results of various experiments indicated that this code could accurately evaluate the annular oxide fuel performance during irradiation in fast reactors. In addition, in order to improve fast reactor performance, optimization of the design margins is required. Consequently, a probabilistic method for fuel rod design has been considered, and the BORNFREE code, which includes the fuel design code as deterministic models to compute the fuel rod behavior and computes the statistical responses of several performance parameters concerned with fuel rod integrity by taking into account the uncertainty of each design parameter, has been developed. The probability that the performance parameters exceed the respective criterion and the design margins can be quantitatively estimated by using BORNFREE. CEPTAR was applied in the deterministic models in BORNFREE, and this system is called BORNFREE-CEPTAR. The improvement of reactor performance for application of the probabilistic design method was studied through probabilistic estimations with the BORNFREE-CEPTAR system. Each design parameter, i.e. dimensions, material and fuel properties, operation conditions, and analysis models, was established as a probability density function based on its statistical characteristics, and the statistical responses of fuel performances were estimated with the Monte Carlo method. The results in this study indicated that the conservative design margins and too rigid specifications as the result of the current deterministic design method could be relationally improved by applying the probabilistic design

  15. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  16. Finite-element procedure for calculating the three-dimensional inelastic bowing of fuel rods (AWBA development program)

    Energy Technology Data Exchange (ETDEWEB)

    Martin, S E

    1982-05-01

    An incremental finite element procedure is developed for calculating the in-pile lateral bowing of nuclear fuel rods. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the results of an axisymmetric stress analysis of the fuel rod. The effects which are taken into account in calculating the rod's lateral bowing include: (a) lateral, axial, and rotational motions and forces at the rod supports, (b) transverse gradients of temperature, fast-neutron flux, and fissioning rate, and (c) cladding circumferential wall thickness variation. The procedure developed in this report could be used to form the basis for a computer program to calculate the time-dependent bowing as a function of the fuel rod's operational and environmental history.

  17. Preliminary design and manufacturing feasibility study for a machined Zircaloy triangular pitch fuel rod support system (grids) (AWBA development program)

    International Nuclear Information System (INIS)

    General design features and manufacturing operations for a high precision machined Zircaloy fuel rod support grid intended for use in advanced light water prebreeder or breeder reactor designs are described. The grid system consists of a Zircaloy main body with fuel rod and guide tube cells machined using wire EDM, a separate AM-350 stainless steel insert spring which fits into a full length T-slot in each fuel rod cell, and a thin (0.025'' or 0.040'' thick) wire EDM machined Zircaloy coverplate laser welded to each side of the grid body to retain the insert springs. The fuel rods are placed in a triangular pitch array with a tight rod-to-rod spacing of 0.063 inch nominal. Two dimples are positioned at the mid-thickness of the grid (single level) with a 900 included angle. Data is provided on the effectiveness of the manufacturing operations chosen for grid machining and assembly

  18. Volatile fission product distributions in LWR spent fuel rods

    International Nuclear Information System (INIS)

    Results presented are from spent fuel characterizations being conducted by the Materials Characterization Center at Pacific Northwest Laboratory on a variety of spent commercial power reactor fuels designated as approved testing materials (ATMs). These ATMs have a variety of burnup levels and fission gas releases; they include fuel from both pressurized water and boiling water reactor designs. The purpose of this work is to provide a source of well-characterized spent fuel for testing in the U.S. Department of Energy Office of Civilian Radioactive Waste Management repository programs and, potentially, other programs

  19. A comparison between genetic algorithms and neural networks for optimizing fuel recharges in BWR

    International Nuclear Information System (INIS)

    In this work the results of a genetic algorithm (AG) and a neural recurrent multi state network (RNRME) for optimizing the fuel reload of 5 cycles of the Laguna Verde nuclear power plant (CNLV) are presented. The fuel reload obtained by both methods are compared and it was observed that the RNRME creates better fuel distributions that the AG. Moreover a comparison of the utility for using one or another one techniques is make. (Author)

  20. A methodology for obtaining the control rod patterns in a BWR using genetic algorithms; Una metodologia para obtener los patrones de barras de control en un BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Montes T, J.L. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Requena R, I. [Universidad de Granada, 18071 Granada (Spain)]. e-mail: jjortiz@nuclear.inin.mx

    2003-07-01

    In this work the GACRP system based on the genetic algorithms technique for the obtaining of the drivers of control bars in a BWR reactor is presented. This methodology was applied to a transition cycle and a one of balance of the Laguna Verde nuclear power station (CNLV). For each one of the studied cycles, it was executed the methodology with a fixed length of the cycle and it was compared the effective multiplication factor of neutrons at the end of the cycle that it is obtained with the proposed drivers of control bars and the multiplication factor of neutrons obtained by means of a Haling calculation. It was found that it is possible to extend several days the length of both cycles with regard to the one Haling calculation. (Author)

  1. Post-irradiation examination of a bowed PWR fuel rod with contact

    International Nuclear Information System (INIS)

    During reactor operation in the Ringhals 2 PWR a fuel rod bowed and as a result came in contact with an adjacent rod. The rod contact in one of the lower grid spans was observed during visual inspection at the end of life. The visual appearance suggested that there was possibly increased cladding corrosion on both the contacting rods at and close to the position of contact. One of the contacting rods was sent to Studsvik’s Hot Cell Laboratory for investigation where fission gas analysis, gamma scanning, EC oxide thickness, metallography (optical microscopy) and cladding microhardness measurements were performed in order to verify the impact of the bow and the contact on the fuel rod performance, with particular focus on the local cladding corrosion. The influence of the reduction of moderator in the region of the contact point was seen in the Cs-137 axial gamma scanning and in the Ce-144 rotational gamma scanning, which show a local reduction of both the pellet-average power, in the contact region, and specifically on the side with the contact. Visual inspection revealed increased corrosion in the rod-rod contact position. Metallographic examination of a cross-section at the elevation with the contact showed that increased corrosion and loss of material had occurred at the contact position. Outside of the immediate vicinity of the contact region the corrosion was not affected. The cladding microhardness was measured at different radial positions both at the contact position and at other positions around the cladding circumference. Based on the relationship between the microhardness and local temperature during operation on fully wetted cladding, it was possible to estimate the cladding surface temperature at the contact point to approximately 360°C. This local overheating and conditions arising from the local overheating can explain the higher local oxidation of the cladding observed in the visual inspection and metallography. (author)

  2. Investigation of Minimum Film boiling Phenomena on Fuel Rods Under Blowdown Cooling Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Stephen M. Bajorek; Michael Gawron; Timothy Etzel; Lucas Peterson

    2003-06-30

    Blowdon cooling heat transfer is an important process that occurs early in a hypothetical large break loss-of-coolant accident (LOCA) in a pressurized water reactor. During blowdown, the flow through the hot assembly is a post-critical heat flux dispersed droplet flow. The heat transfer mechanisms that occur in blowdown cooling are complex and depend on droplet and heated surface interaction. In a safety analysis, it is of considerable importance to determine the thermal-hydraulic conditions leading to the minimum film boiling temperature, Tmin. A flow boiling rig for measurement of blowdown cooling heat transfer and quench phenomena on a nuclear fuel rod simulator was designed and constructed for operation at up to 12.4 MPa. The test section consisted of a concentric annulus, with a 9.5 mm OD nuclear fuel rod simulator at the center. The rod was contained within a 0.85 mm thick, 19 mm OD 316 stainless steel tube, forming the flow channel. Two types of rods were tested; one type was sheathed with Inconel 600 while the other was clad with Zircaloy-2. Water was injected into the test section at the top of the heated length through an injection header. This header was an annular sign that fit around the fuel rod simulator and within the stainless steel tube. Small spacers aligned the injection header and prevented contract with either the heater rod or the tube. A series of small diameter holes at the bottom of the header caused the formation of droplets that became entrained with the steam flow. The test section design was such that quench would take place on the rod, and not along the channel outer annulus.

  3. Prediction of the local power factor in BWR fuel cells by means of a multilayer neural network; Prediccion del factor local de potencia en celdas de combustible BWR mediante una red neuronal multicapas

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.L.; Ortiz, J.J.; Perusquia C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 La Marquesa, Ocoyoacac, Estado de Mexico (Mexico); Francois, J.L.; Martin del Campo M, C. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: jlmt@nuclear.inin.mx

    2007-07-01

    To the beginning of a new operation cycle in a BWR reactor the reactivity of this it increases by means of the introduction of fresh fuel, the one denominated reload fuel. The problem of the definition of the characteristics of this reload fuel represents a combinatory optimization problem that requires significantly a great quantity of CPU time for their determination. This situation has motivated to study the possibility to substitute the Helios code, the one which is used to generate the new cells of the reload fuel parameters, by an artificial neuronal network, with the purpose of predicting the parameters of the fuel reload cell of a BWR reactor. In this work the results of the one training of a multilayer neuronal net that can predict the local power factor (LPPF) in such fuel cells are presented. The prediction of the LPPF is carried out in those condition of beginning of the life of the cell (0.0 MWD/T, to 40% of holes in the one moderator, temperature of 793 K in the fuel and a moderator temperature of 560 K. The cells considered in the present study consist of an arrangement of 10x10 bars, of those which 92 contains U{sup 235}, some of these bars also contain a concentration of Gd{sub 2}O{sub 3} and 8 of them contain only water. The axial location inside the one assembles of recharge of these cells it is exactly up of the cells that contain natural uranium in the base of the reactor core. The training of the neuronal net is carried out by means of a retro-propagation algorithm that uses a space of training formed starting from previous evaluations of cells by means of the Helios code. They are also presented the results of the application of the neuronal net found for the prediction of the LPPF of some cells used in the real operation of the Unit One of the Laguna Verde Nuclear Power station. (Author)

  4. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    Energy Technology Data Exchange (ETDEWEB)

    Kee, E. [Houston Lighting & Power Co., Wadworth, TX (United States)

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  5. Power ramp tests of MOX fuel rods. HBWR irradiation with the instrument rig, IFA-591

    International Nuclear Information System (INIS)

    Plutonium-uranium mixed oxide (MOX) fuel rods of instrumental rig IFA-591 were ramped in HBWR to study the Advanced Thermal Reactor (ATR) MOX fuel behavior during transient operation and to determine a failure threshold of the MOX fuel rods. Eleven segments were base-irradiated in ATR 'FUGEN' up to 18.4 GWd/t. Zirconium liner claddings were adopted for four segments of them. As the results of non-destructive post irradiation examinations (PIEs) after the base-irradiation and before the ramp tests, no remarkable behavior affecting the integrity of fuel assembly and fuel rod was confirmed. All segments to be used for the ramp tests, which consisted of the multi-step ramp tests and the single-step ramp tests, had instrumentations for in-pile measurements of cladding elongation or plenum pressure, and heated up to the maximum linear power of 58.3-68.4 kW/m without failure. The major results of ramp tests are as follows: There is no difference in PCMI behaviors between two type rods of Zry-2 and Zirconium liner claddings from the in-pile measurements of cladding elongation and plenum pressure. The computations of cladding elongation and inner pressure gave slightly lower elongation and pressure than the in-pile measurements during the ramp-test. However, the cladding relaxation during the power hold was in good agreement, and the fission gas release behavior during cooling down could be evaluated by taking into account the relaxation of contact pressure between pellet and cladding. Although the final power during IFA-591 ramp tests reached the higher linear power than the failure threshold power of UO2 fuel rods, no indication of fuel failure was observed during the ramp tests. The cladding relaxation due to the creep deformation of the MOX pellets at high temperature could be confirmed at the power steps during the multi-ramp test. The fission gas release due to the emancipation from PCMI stress was observed during the power decreasing. The burn-up dependence could be

  6. Prototypical spent fuel rod consolidation equipment preliminary design report: Volume 1, Report

    International Nuclear Information System (INIS)

    This design report describes the NUS Preliminary Design of the Prototype Spent Nuclear Fuel Rod Consolidation Equipment for the Department of Energy. The sections of the report elaborate on each facet of the preliminary design. A concept summary is provided to assist the reader in rapidly understanding the complete design. The NUS Prototype Spent Fuel Rod Consolidation System is an automatically controlled system to consolidate a minimum of 750 MT (heavy metal)/year of US commercial nuclear reactor fuel, at 75% availability. The system is designed with replaceable components utilizing the latest state-of-the-art technology. This approach gives the system the flexibility to be developed without costly development programs, yet accept new technology as it evolves over the next ten years. Capability is also provided in the system design to accommodate a wide variety of fuel conditions and to recover from any situation which may arise

  7. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  8. A methodology for obtaining the control rods patterns in a BWR using systems based on ants colonies; Una metodologia para obtener los patrones de barras de control en un BWR usando sistemas basados en colonias de hormigas

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J. [Depto. de Sistemas Nucleares, ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Requena R, I. [Universidad de Granada, 18071 Granada (Spain)]. e-mail: jjortiz@nuclear.inin.mx

    2003-07-01

    In this work the AZCATL-PBC system based on a technique of ants colonies for the search of control rods patterns of those reactors of the Nuclear Power station of Laguna Verde (CNLV) is presented. The technique was applied to a transition cycle and one of balance. For both cycles they were compared the k{sub ef} values obtained with a Haling calculation and the control rods pattern proposed by AZCATL-PBC for a burnt one fixed. It was found that the methodology is able to extend the length of the cycle with respect to the Haling prediction, maintaining sure to the reactor. (Author)

  9. Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

    International Nuclear Information System (INIS)

    The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

  10. Improvement of input parameters for the estimation of fuel rod temperature in dry transport cask

    International Nuclear Information System (INIS)

    A typical PWR spent fuel bundle has a 17 x 17 rod array, and an analysis requires a very long computation time and a vast amount of memory. Therefore, we applied the lumped fuel bundle analysis approach with the homogenized method to estimate the fuel cladding temperature efficiency. Thermal analysis results for lumped fuel bundles showed an excessive radiative heat transfer, and we applied an emissivity modification factor to compensate for this radiation effect. The value of the factor decreased as the number of the rods in the homogenized array decreased.. For the lumped 8 x 8 array, the best emissivity modification factor was shown to be 0.40. The rod emissivity of 0.8 is generally recommended to be used in COBRA-SFS[D. R. Rector et al.] calculations. Therefore, we can use the modified rod emissivity of 0.32 for lumped 8 x 8 array. There are good agreements between the results from lumped 8 x 8 array bundle and the results from real 17 x 17 array bundle. By homogenization, we can increase the computational speed substantially, as well as reduce the requirements on computer memory and space. (authors)

  11. Design of Testing Set-up for Nuclear Fuel Rod by Neutron Radiography at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HAO; Li-jie; WU; Mei-mei; HE; Lin-feng; WANG; Yu; LIU; Yun-tao; SUN; Kai; CHEN; Dong-feng

    2012-01-01

    <正>An experimental set-up dedicated to non-destructively test a 15 cm long pressurized water reactor (PWR) nuclear fuel rod by neutron radiography (NR) is designed and fabricated. It consists of three parts: Transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo simulation by the MCNP code.

  12. Prototypical spent fuel rod consolidation equipment preliminary design report: Volume 2, Drawings

    International Nuclear Information System (INIS)

    This volume consists of 65 E size drawings and 4 sketches of the NUS spent fuel rod consolidation equipment. The drawings have been grouped into categories; a detailed list of the drawings is included. The sketches prepared during the preliminary design process have been included

  13. NDT of the fuel rods with artificial defect

    Energy Technology Data Exchange (ETDEWEB)

    Yang, S.Y.; Min, D.K.; Eom, S.H.; Chun, Y.B.; Min, D.K

    2000-07-01

    Non-destructive examination such as visual inspection, dimensional measurement, eddy current and gamma scanning have been carried out. The objective of this study is to evaluate the characteristics of spent fuels, and to obtain the basic technical data through the study of long term storage behavior of spent fuels. In the results of visual inspection, there is no observable effects around the part of artificial defect. And there is nothing unusual in the results of gamma scanning. Diameter and ovality the artificial defect were measured. The result obtained from this study will be used as a basic data for the study of behavior for spent fuel under the long term storage condition and the safety evaluation of spent fuel.

  14. Volatile fission product distributions in LWR fuel rods

    International Nuclear Information System (INIS)

    Results from this study are a part of spent fuel characterizations being conducted by the Materials Characterization Center (MCC) project at Pacific Northwest Laboratory on a variety of spent uranium oxide fuels designated as Approved Testing Materials (ATMs). These ATMs have a variety of burnup levels, fission gas releases, and include fuel from both pressurized water and boiling water reactor designs. The purpose of this work is to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) repository programs and potentially other programs. Details of these characterization studies for some of the ATMs are available. 7 refs., 8 figs., 1 tab

  15. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  16. Qualification and interpretation of MR test reactor irradiation data on WWER-440 type fuel rods for fuel thermal model validation

    International Nuclear Information System (INIS)

    Conducted jointly by the Kurchatov Institute of Atomic Energy of the USSR and the Finnish state owned power utility Imatran Voima Oy, an irradiation program, SOFIT, including a series of characterised and instrumented test rod bundles is in progress to examine the thermal and mechanical behaviour of WWER type fuel rods. The main purposes of the work are to promote the flexibility and the economy of the fuel utilisation, and to verify, for the WWER annular fuel and Zr1%Nb cladding, the fuel behaviour computer programs used in the Soviet Union and in Finland. The more specific objectives are to obtain information on fuel temperature, internal gas pressure and cladding elongation during irradiation for different fabrication parameters. Qualification and interpretation of in-pile results from the first irradiation, completed at maximum local exposure of about 16 MWd/kgU, are presented. Quantitative evaluation of affecting phenomena could be deduced from the restored thermocouple readings. The measured temperatures show dependency on the as-fabricated gap size and weakness on the helium fill pressure. The results of the calculations, obtained by the existing steady state fuel behaviour computer codes are compared with the experimental temperatures. The importance of careful qualification of the data is emphasised. The related uncertainties are discussed. Preliminary fission gas data of some of the rods have become available, which show low or moderate gas release. (author). 5 refs, 7 figs, 1 tab

  17. Studies of a larger fuel bundle for the ABWR improved evolutionary reactor

    International Nuclear Information System (INIS)

    Studies for an Improved Evolutionary Reactor (IER) based on the Advanced Boiling Water Reactor (ABWR) were initiated in 1990. The author summarizes the current status of the core and fuel design. A core and fuel design based on a BWR K-lattice fuel bundle with a pitch larger than the conventional BWR fuel bundle pitch is under investigation. The core and fuel design has potential for improved core design flexibility and improved reactor transient response. Furthermore, the large fuel bundle, coupled with a functional control rod layout, can achieve improvement of operation and maintenance, as well as improvement of overall plant economy

  18. Method of handling and/or storing a nuclear fuel assembly consisting of an elongated frame that contains fuel rods and fuel assembly designed specially for this method

    International Nuclear Information System (INIS)

    In order to assure subcriticality during handling, transport and/ore storage of nuclear reactor fuel assemblies and additional body containing a neutron absorbing material and touching beside the fuel rods is fixed to the frame of the fuel assembly. This body has a handle with an adapted coupling element mounted on a holding device for handling and/or storage of the fuel assembly. (orig./RW)

  19. Thermocouples recovery of one TRIGA instrumented fuel rod

    International Nuclear Information System (INIS)

    The IPR-R1 TRIGA Mark-I research reactor at the Nuclear Technology Development Centre - CDTN (Belo Horizonte/Brazil) reached its first criticality on November 1960, with a core configuration containing 56 aluminum clad standard TRIGA fuel elements and a maximum thermal power of 30 kW. In order to upgrade the IPR-R1 reactor power, nine stainless steel clad fuel elements were purchased in 1971. One of these fuel elements was instrumented in the centreline with three type K thermocouples. On December 2000, four of these stainless steel clad fuel elements were placed into the core allowing to upgrading the nominal power to 250 kW. In 2004 the instrumented fuel element (IF) was inserted into the core hottest position, predicted by neutronic calculations. The IF stayed in this position up to 2007, allowing heat transfer investigations in several operating powers, including the maximum power of 250 kW. During this time it also monitored the core temperature in all operations. The fuel temperature and other operational parameters were stored in a computer hard disk, with an accessible historical database, in order to make the chronological information on reactor performance and its behavior available to users. After almost 3 years of monitoring the core temperature, it was noticed that the three thermocouples failed in its measures. It was observed the rupture of the thermocouples in the connector placed between the thermo-elements wire and its extension cables. This paper describes the methodology used in the recovery of the instrumented fuel element thermocouples, carried out in October of 2008 at CDTN. It was obtained partial success in the restoration of the thermo-elements continuity. There are suggested procedures for new recovery and the returning of the IF to the core, improving the operational safety to the reactor. According to the specification of General Atomics Electronic Systems Inc instrumentation, and like several TRIGA reactors in operation today, it is

  20. Whole-rod testing of intact and defective LWR rods under expected dry-storage conditions

    International Nuclear Information System (INIS)

    The objective of this project is to provide the Nuclear Regulatory Commission with information to confirm or establish spent fuel dry storage licensing positions relative to: (1) the long-term, low-temperature (less than 2500C) behavior of spent fuel rods in dry storage; and (2) the radioactive contamination potential of crud from cladding in dry storage. The basic need for this data is to: confirm long-term, low temperature (less than 2500C) spent fuel dry storage performance predictions based on theoretical analyses and on results from high-temperature, short-term laboratory tests; determine the nature and behavior of crud layers as a function of dry storage time; and determine the potential radioactive crud contamination (e.g., spalling characteristics) for dry storage. An eight-rod test matrix of PWR and BWR rods was chosen which consisted of all combinations of intact or breached cladding in an oxidizing or inert atmosphere. The PWR rods (30.5 GWD/MTU) were discharged from H.B. Robinson in May 1974, and the BWR rods (12.9 GWD/MTU) were discharged from Peach Bottom in March 1976. The eight test rods were visually inspected for crud and defects with the results recorded on video tape. Cladding penetration was confirmed. All the rods were put in test capsules with the appropriate atmosphere and leak checked. The test capsules were loaded into a test train and the train was placed in the furnace cavity. The test was started on September 15, 1982 and is presently at 2300C. After the first 10-month run is completed, an interim examination, consisting of visual inspection, gamma scanning, and crud sampling, will be conducted

  1. Out-of-pile experiments on the high temperature behavior of Zry-4 clad fuel rods

    International Nuclear Information System (INIS)

    Out-of-pile experiments have been performed to investigate the escalation in temperature of zircaloy clad fuel rods during heatup in steam due to the exothermal zircaloy steam reaction. In these tests single zircaloy/UO2 fuel rod simulators surrounded with a zircaloy shroud - simulating the zircaloy of neighbouring rods - were heated inside a fiber ceramic insulation. The initial heating rates were varied from 0.3 to 2.5 K/s. In every test an escalation of the temperature rise rate was observed. The maximum measured surface temperature was about 22000C. The temperature decreased after the maximum had been reached without decreasing the input electric power. The temperature decreases were due to inherent processes including the runoff of molten zircaloy. The escalation process was influenced by the temperature behaviour of the shroud, which was itself affected by the insulation and steam cooling. Damage to the fuel rods increased with increasing heatup rate. For slow heat up rates nearly no interaction between the oxidized cladding and UO2 was observed, while for fast heatup rates the entire annular pellet was dissolved by molten zircaloy. (orig.)

  2. Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

    International Nuclear Information System (INIS)

    The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B4C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B4C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B4C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.)

  3. Equipment and methods for examinations of fuel rods in the MIR reactor storage pool

    International Nuclear Information System (INIS)

    A wide range of tests of fuel rods and structural materials of water-cooled power reactors is performed in the loop facilities (LFs) of the MIR reactor. Depending on the objectives and tasks of different experiments, the performance of periodical interim examinations of irradiated items is required. However, as a result of some circumstances, it is not always possible to conduct them in hot cells. In this context, JSC 'SSC RIAR' has developed the equipment for interim examinations of fuel rods and design components of the experimental fuel assemblies (EFAs) in the MIR reactor storage pool (SP). Besides, this equipment can be used for cleaning of the examined items from surface deposits prior to the measurements. The paper describes the main characteristics and capabilities of the developed equipment, methodical aspects of the performed interim examinations, as well as some experimental results obtained using this equipment. In future, its upgrade is planned. (author)

  4. Experience in generalizing the data on the rod fuel assembly burnout by the method of cells

    International Nuclear Information System (INIS)

    The comparative analysis of results of calculations on the rod fuel assembly burnout by the method of cells is conducted. The method of cells is based on determining local parameters of coolant flow in the assembly cross section. The channel cross section is conditionally subdivided into elementary cells communicating with each other. For such system of interacting cells the equations of thermohydraulics are solved. Local values of enthalpy and mass rate obtained are used for fuel burnout calculation. The analysis performed has shown that generalizing the empirical data for a given set of experiments on the basis of local parameters of flow in the cells already nowadays assures in most cases accuracy of burnout conditions prediction which is not less than the traditional dependences based on one-dimensional description of the flow. The conclusion is drawn on the prospects of using the method of cells for calculating the burnout in rod fuel assemblies and necessity of its subsequent development

  5. Development and validation of the FEM based global Fuel Rod Analyzer FRA-TF-global

    International Nuclear Information System (INIS)

    Poster summarizes results and experience gained during development of global FRA-TF fuel rod FEM simulator. Step by step is explained design of an AutoCAD 3D model of the WWER-440 typical fuel rod and WWER-1000 typical fuel rod. All the inner fuel rod design parts are modeled and connected in the whole model - object. The transfer of prepared 3D models into the COSMOSDesignSTAR/COSMOSM system was made and the various types of meshing as well as solid type of FEM elements tested. Fuel rod is modeled with all inner design parts appropriately meshed for the thermophysical simulation (thermal field with all thermal sources and heat local and global transfers). Example of the software limitations is given. The FRATFM versions (WWER- 440 and WWER-1000) are using nonlinear thermal solver (HSTAR), static mechanics solver (STAR) and nonlinear (NSTAR) mechanics COSMOSM solver. All materials, thermophysical properties as well as mechanical properties are prepared in the form of COSMOSM library - functions and material curves with temperature dependence and time (expressing burn-up process). Successful implementation of a simulator for the of-line usage at the NPP and its validation against the FEM based integral code FEMAXI is given. Open questions of this type of development and industrial implementation are discussed in the conclusions. Computer techniques and practical applicability is critically commented using real case of a 5 year irradiation history calculation of a typical WWER-1000 fuel rod. This poster closes one phase of our work, which started with classical safety documentation and 2D sketches as input and ended in the modern AutoCADCOSMOSDesignSTAR 3D application delivered to the industry customer. Developed models can be transferred to another FEM based program that has more features. References: Belytschko, T., Liu, W. K., Moran, B.:Nonlinear Finite Elements for Continua and Structures. J. Wiley and Sons Ltd., 2000. Hughes,T.J.R.: The Finite Element Method

  6. Information to be requested from the NSSS vendor for fuel management capability for BWR

    Energy Technology Data Exchange (ETDEWEB)

    Minguez, E.; Esteban, A.; Gomez, M.; Leira, G.; Martinez, R.; Serrano, J.

    1975-07-01

    A set of the nuclear, thermal-hydraulic, and mechanical parameters necessary according to the design of BWRs, is listed. This parameters are necessary to perform the fuel elements management and design, and it must be supplied by the Reactor Manufacturer to the Utility. (Author) 18 refs.

  7. Characterisation of high-burnup LWR fuel rods through gamma tomography

    International Nuclear Information System (INIS)

    Current fuel management strategies for light water reactors (LWRs), in countries with high back-end costs, progressively extend the discharge burnup at the expense of increasing the 235U enrichment of the fresh UO2 fuel loaded. In this perspective, standard non-destructive assay techniques, which are very attractive because they are fast, cheap, and preserve the fuel integrity, in contrast to destructive approaches, require further validation when burnup values become higher than 50 GWd/t. This doctoral work has been devoted to the development and optimisation of non-destructive assay techniques based on gamma-ray emissions from irradiated fuel. It represents an important extension of the unique, high-burnup related database, generated in the framework of the LWR PROTEUS Phase II experiments. A novel tomographic measurement station has been designed and developed for the investigation of irradiated fuel rod segments. A unique feature of the station is that it allows both gamma-ray transmission and emission computerised tomography to be performed on single fuel rods. Four burnt UO2 fuel rod segments of 400 mm length have been investigated, two with very high (52 GWd/t and 71 GWd/t) and two with ultra-high (91 GWd/t and 126 GWd/t) burnup. Several research areas have been addressed, as described below. The application of transmission tomography to spent fuel rods has been a major task, because of difficulties of implementation and the uniqueness of the experiments. The main achievements, in this context, have been the determination of fuel rod average material density (a linear relationship between density and burnup was established), fuel rod linear attenuation coefficient distribution (for use in emission tomography), and fuel rod material density distribution. The non-destructive technique of emission computerised tomography (CT) has been applied to the very high and ultra-high burnup fuel rod samples for determining their within-rod distributions of caesium and

  8. Optimized design of the nuclear fuel rod transport container used for non-destructive testing with neutron radiography

    International Nuclear Information System (INIS)

    Background: Working under extreme conditions, nuclear fuel rods, the key component of nuclear plants and reactors, are easy to be broken. In order to be safe in operation, lots of testing methods on the fuel rods have to be carried out from fabrication to operation. Purpose: Neutron radiography is a unique non-destructive testing technique which can be used to test samples with radioactivity. As the essential equipment, the nuclear fuel rod transport container has to shield the radioactivity of fuel rod and control its movement during testing and transporting. Methods: The shielding simulation of the transport container was performed using the MCNP4C code, which is a general purpose Monte Carlo code for calculating the time dependent multi-group energy transport equation for neutrons, photons and electrons in three dimensional geometries. Results: The material and dimension of the transport container used for neutron radiography testing fuel rods at Chinese Advanced Research Reactor (CARR) were optimally designed by MCNP, and the mechanical devices used to control fuel rods' movement were also described. Conclusion: The 2-m long fuel rod can be tested at CARR's neutron radiography facility (under construction) with this transport container. (authors)

  9. Sturdy on Orbital TIG Welding Properties for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Hong, Jintae; Kim, Kahye; Huh, Sungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    We developed a precision TIG welding system that is able to weld the seam between end-caps and a fuel cladding tube for the nuclear fuel test rod and rig. This system can be mainly classified into an orbital TIG welder (AMI, M-207A) and a pressure chamber. The orbital TIG welder can be independently used, and it consists of a power supply unit, a microprocessor, water cooling unit, a gas supply unit and an orbital weld head. In this welder, the power supply unit mainly supplies GTAW power for a welding specimen and controls an arc starting of high frequency, supping of purge gas, arc rotation through the orbital TIG welding head, and automatic timing functions. In addition, the pressure chamber is used to make the welded surface of the cladding specimen clean with the inert gas filled inside the chamber. To precisely weld the cladding tube, a welding process needs to establish a schedule program for an orbital TIG welding. Therefore, the weld tests were performed on a cladding tube and dummy rods under various conditions. This paper describes not only test results on parameters of the purge gas flow rates and the chamber gas pressures for the orbital TIG welding, but also test results on the program establishment of an orbital TIG welding system to weld the fuel test rods. Various welding tests were performed to develop the orbital TIG welding techniques for the nuclear fuel test rod. The width of HAZ of a cladding specimen welded with the identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a weld end-point because of heat accumulation. The welding effect of the PGFR and CGP shows a relatively large difference for FSS and LSS. Each hole on the cladding specimens was formed in the 1bar CGP with the 20L/min PGFR but not made in the case of the PGFR of 10L/min in the CGP of 2bar. The optimum schedule program of the orbital TIG welding system to weld the nuclear fuel test rod was established through the program

  10. Sturdy on Orbital TIG Welding Properties for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    We developed a precision TIG welding system that is able to weld the seam between end-caps and a fuel cladding tube for the nuclear fuel test rod and rig. This system can be mainly classified into an orbital TIG welder (AMI, M-207A) and a pressure chamber. The orbital TIG welder can be independently used, and it consists of a power supply unit, a microprocessor, water cooling unit, a gas supply unit and an orbital weld head. In this welder, the power supply unit mainly supplies GTAW power for a welding specimen and controls an arc starting of high frequency, supping of purge gas, arc rotation through the orbital TIG welding head, and automatic timing functions. In addition, the pressure chamber is used to make the welded surface of the cladding specimen clean with the inert gas filled inside the chamber. To precisely weld the cladding tube, a welding process needs to establish a schedule program for an orbital TIG welding. Therefore, the weld tests were performed on a cladding tube and dummy rods under various conditions. This paper describes not only test results on parameters of the purge gas flow rates and the chamber gas pressures for the orbital TIG welding, but also test results on the program establishment of an orbital TIG welding system to weld the fuel test rods. Various welding tests were performed to develop the orbital TIG welding techniques for the nuclear fuel test rod. The width of HAZ of a cladding specimen welded with the identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a weld end-point because of heat accumulation. The welding effect of the PGFR and CGP shows a relatively large difference for FSS and LSS. Each hole on the cladding specimens was formed in the 1bar CGP with the 20L/min PGFR but not made in the case of the PGFR of 10L/min in the CGP of 2bar. The optimum schedule program of the orbital TIG welding system to weld the nuclear fuel test rod was established through the program

  11. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  12. Serpent: an alternative for the nuclear fuel cells analysis of a BWR

    International Nuclear Information System (INIS)

    In the last ten years the diverse research groups in nuclear engineering of the Universidad Nacional Autonoma de Mexico and Instituto Politecnico Nacional (UNAM, IPN), as of research (Instituto Nacional de Investigaciones Nucleares, ININ) as well as the personnel of the Nuclear Plant Management of the Comision Federal de Electricidad have been using the codes Helios and /or CASMO-4 in the generation of cross sections (X S) of nuclear fuel cells of the cores corresponding to the Units 1 and 2 of the nuclear power plant of Laguna Verde. Both codes belong to the Studsvik-Scandpower Company who receives the payment for the use and their respective maintenance. In recent years, the code Serpent appears among the nuclear community distributed by the OECD/Nea which does not has cost neither in its use neither in its maintenance. The code is based on the Monte Carlo method and makes use of the processing in parallel. In the Escuela Superior de Fisica y Matematicas of the IPN, the personnel has accumulated certain experience in the use of Serpent under the direction of personal of the ININ; of this experience have been obtained for diverse fuel burned, the infinite multiplication factor for three cells of nuclear fuel, without control bar and with control bar for a known thermodynamic state fixed by: a) the fuel temperature (Tf), b) the moderator temperature (Tm) and c) the vacuums fraction (α). Although was not realized any comparison with the X S that the codes Helios and CASMO-4 generate, the results obtained for the infinite multiplication factor show the prospective tendencies with regard to the fuel burned so much in the case in that is not present the control bar like when it is. The results are encouraging and motivate to the study group to continue with the X S generation of a core in order to build the respective library of nuclear data as a following step and this can be used for the codes PARCS, of USA NRC, DYN3D of HZDR, or others developed locally in the IPN

  13. Demonstration test of the spent fuel rod cutting process with tube cutter mechanism

    International Nuclear Information System (INIS)

    In this paper, the verification by computer graphics technology for the spent fuel rod cutting devise which belongs to the spent fuel disassembly processes, the performance tests of the real device, and the demonstration tests with tube cutter mechanism are described. The graphical design system is used throughout the design stages from conceptual design to motion analysis like collision detection. By using this system, the device and the process are optimized. The performance test of the real device and the demonstration test using the tube cutter mechanism in the hot cell are carried out. From these results, the spent fuel rod cutting device is improved based on the considerations of circularity of the rod cross-section, debris generation, and fire risk etc. Also, this device is improved to be operated automatically via remote control system considering later use in closed environment like Hot-cell (radioactive area) and the modulization in the structure of this device makes maintenance easy. The result of the performance test and the demonstration in this report is expected to contribute to the optimization of the pre-treatment processes for the reuse of the spent fuel like DUPIC process and the final disposal

  14. Comparison of heuristic optimization techniques for the enrichment and gadolinia distribution in BWR fuel lattices and decision analysis

    International Nuclear Information System (INIS)

    Highlights: • Different metaheuristic optimization techniques were compared. • The optimal enrichment and gadolinia distribution in a BWR fuel lattice was studied. • A decision making tool based on the Position Vector of Minimum Regret was applied. • Similar results were found for the different optimization techniques. - Abstract: In the present study a comparison of the performance of five heuristic techniques for optimization of combinatorial problems is shown. The techniques are: Ant Colony System, Artificial Neural Networks, Genetic Algorithms, Greedy Search and a hybrid of Path Relinking and Scatter Search. They were applied to obtain an “optimal” enrichment and gadolinia distribution in a fuel lattice of a boiling water reactor. All techniques used the same objective function for qualifying the different distributions created during the optimization process as well as the same initial conditions and restrictions. The parameters included in the objective function are the k-infinite multiplication factor, the maximum local power peaking factor, the average enrichment and the average gadolinia concentration of the lattice. The CASMO-4 code was used to obtain the neutronic parameters. The criteria for qualifying the optimization techniques include also the evaluation of the best lattice with burnup and the number of evaluations of the objective function needed to obtain the best solution. In conclusion all techniques obtain similar results, but there are methods that found better solutions faster than others. A decision analysis tool based on the Position Vector of Minimum Regret was applied to aggregate the criteria in order to rank the solutions according to three functions: neutronic grade at 0 burnup, neutronic grade with burnup and global cost which aggregates the computing time in the decision. According to the results Greedy Search found the best lattice in terms of the neutronic grade at 0 burnup and also with burnup. However, Greedy Search is

  15. Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod

    International Nuclear Information System (INIS)

    In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR's operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends

  16. Fuel rod internal chemistry and fission products behaviour

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the members of the International Working Group on Water Reactor Fuel Performance and Technology. Forty-six participants representing fourteen countries and two international organizations attended the meeting. Twenty-one presentations were discussed in four sessions: thermodynamics of fission products (six papers); fission products migration and release (seven papers); fission product release in transients or accident conditions (four papers); fission products to cladding interaction - stress corrosion cracking (five papers). A separate abstract was prepared for all twenty-one papers

  17. Completion of UO2 pellets production and fuel rods load for the RA-8 critical facility

    International Nuclear Information System (INIS)

    The Advanced Fuels Division produced fuel pellets of 235U with 1.8% and 3.6% enrichment and Zry-4 cladding loads for the RA-8 reactor at Pilcaniyeu Technological Unit. For economical and availability reasons, the powder acquired was initially UO2 with 3.4% enrichment in 235U, therefore the 235U powder with 1.8% enrichment was produced by mechanical mixture. The production of fuel pellets for both enrichments was carried out by cold pressing and sintering processes in reducing atmosphere. The load of Zry-4 claddings was performed manually. The production stages can be divided into setup, qualification and production. This production allows not only to fulfill satisfactorily the new fuel rods supply for the RA-8 reactor but also to count with a new equipment and skilled personnel as well as to meet quality and assurance control methods for future pilot-scale production and even new fuel elements production. (author)

  18. Fission gas release and fuel rod chemistry related to extended burnup

    International Nuclear Information System (INIS)

    The purpose of the meeting was to review the state of the art in fission gas release and fuel rod chemistry related to extended burnup. The meeting was held in a time when national and international programmes on water reactor fuel irradiated in experimental reactors were still ongoing or had reached their conclusion, and when lead test assemblies had reached high burnup in power reactors and been examined. At the same time, several out-of-pile experiments on high burnup fuel or with simulated fuel were being carried out. As a result, significant progress has been registered since the last meeting, particularly in the evaluation of fuel temperature, the degradation of the global thermal conductivity with burnup and in the understanding of the impact on fission gas release. Fifty five participants from 16 countries and one international organization attended the meeting. 28 papers were presented. A separate abstract was prepared for each of the papers. Refs, figs, tabs and photos

  19. In-pile experiments on the fuel rod behavior in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    This report describes the results of the destructive postirradiation examination of Test Series F with respect to the mechanical behavior of Zircaloy-4 cladding material, the oxidation and corrosion behavior of the cladding, the mechanical behavior of the fuel, and the fission gas release and volume change of the fuel. No difference between the behavior of unirradiated and preirradiated rods has been found, with the exception of the fuel condition. In the preirradiated fuel rods the fuel underwent considerable fragmentation during the experiments. Oxidation of the inside surface of the cladding was observed at and near the rupture and was caused primarily by penetrating steam. The fuel contributes slightly to oxidation of the cladding inside surface. No influence of iodine or other volatile fission products on the deformation and fracture behavior due to stress corrosion cracking has been found. The maximum fission gas release during preirradiation was 5.5% with an additional 6% release during the in-pile LOCA transient. (orig./HP)

  20. Thermal simulation experiments with a UO2 fuel rod assembly

    International Nuclear Information System (INIS)

    Evidence is presented which shows that columnar grains can be induced to grow in high-density sintered uranium dioxide specimens by applying a steep temperature gradient at temperatures above 1700oC but below the melting point of 2800oC. Columnar growth apparently is a result of the migration of large transverse voids, whose individual widths define the cross sections of the grains, up a temperature gradient by a sublimation process. The grains grown by this process have a (iii) preferred orientation along their columnar axis. A consequence of such void migration in operating fuel elements containing solid UO2 pellets is the formation of a central void bounded by a region of oxide exhibiting columnar growth. (author)

  1. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1990-02-01

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

  2. Fuel Services Germany - AREVA NP

    Energy Technology Data Exchange (ETDEWEB)

    Langenberger, Jan; Seidel, Jana [AREVA, AREVA NP, P.O Box 11 09, 91001 Erlangen (Germany)

    2008-07-01

    AREVA NP has over 30 years of nuclear fuel inspection and repair experience. Our service staff is trained regularly and holds a long lasting experience in their field of work. Our services comprise all activities within examinations of lead test assemblies, all necessary inspections and tests of fuel assemblies and other core components during an outage, reconstitution of fuel assemblies for further use in the reactor and activities within the disposal of fuel assemblies and core components. Furthermore we deliver facilities and systems that are necessary for the above mentioned services. These are in particular facilities for inspections, sipping, repair and for disposal (e.g. conditioning equipment, facility for encapsulation of leaking fuel rods). Continuous research and development guarantee that the applied facilities always comply with the latest state of the art. They are suitable for a variety of fuel assembly designs of AREVA's and external supplier's latest and conventional technology. The facilities are applicable in different BWR-, PWR and VVER nuclear power plants. Inspections in PWR- and BWR Plants: - Measurements with contact and contact less measuring methods of fuel assemblies, fuel rods, fuel channels and other core components; - High-definition video inspections; - Mast-, In-core- and Box-sipping to detect defective fuel assemblies; - Support (e.g. measurements on site) for post irradiation examinations; - Sampling of material (e.g. taking CRUD or material probes) and fuel rod preparation for hot cell examination. Repair in PWR- and BWR Plants: - Exchange of defective fuel rods of a fuel assembly for further use in the reactor and preparation of the fuel assembly or defective fuel rod for disposal; - Reconstitution of fuel assemblies with damaged structure parts (spacer, springs, upper-and lower end fitting). We offer products that meet the requirements of pressure-tightness under water (e. g. for use in the fuel storage pool), that are

  3. Modelling of pressurized water reactor fuel, rod time dependent radial heat flow with boundary element method

    International Nuclear Information System (INIS)

    The basic principles of the boundary element method numerical treatment of the radial flow heat diffusion equation are presented. The algorithm copes the time dependent Dirichlet and Neumann boundary conditions, temperature dependent material properties and regions from different materials in thermal contact. It is verified on the several analytically obtained test cases. The developed method is used for the modelling of unsteady radial heat flow in pressurized water reactor fuel rod. (author)

  4. Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

    OpenAIRE

    Amir Zacarias Mesquita; Rogério Rivail Rodrigues

    2013-01-01

    In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF). At this time,...

  5. Structure Optimization Design of the Electronically Controlled Fuel Control Rod System in a Diesel Engine

    OpenAIRE

    Hui Jin; Haosen Wang

    2015-01-01

    Poor ride comfort and shorter clutch life span are the key factors restricting the commercialization of automated manual transmission (AMT). For nonelectrically controlled engines or AMT where cooperative control between the engine and the transmission is not realizable, applying electronically controlled fuel control rod systems (ECFCRS) is an effective way to solve these problems. By applying design software such as CATIA, Matlab and Simulink, and MSC Adams, a suite of optimization design m...

  6. Fabrication and loading of fuel rods for the Light Water Breeder Reactor (LWBR Development Program)

    International Nuclear Information System (INIS)

    The fabrication and inspection operations used for the manufacture of approximately 24,000 fuel rods for the Light Water Breeder Reactor are described in detail. This report also describes the development work to establish the fabrication procedures and investigations undertaken to solve problems encountered during manufacturing. The approximately 10 foot long LWBR fuel rods were made in four outside diameters ranging from 0.306 inch (seed) to 0.832 inch (reflector). Each rod was fabricated by sealing cylindrical oxide fuel pellets (ThO2-U233O2), into Zircaloy seamless tubes by welding Zircaloy enclosures at the ends. The special inspections performed to assure a high quality product meeting all design requirements are described. These inspections included weld radiography and ultrasonic inspection, in-motion radiography to evaluate internal dimensions and pellet integrity, helium leak testing, corrosion testing, and detection of surface contamination. The facilities designed and built for this fabrication effort are described and the resultant manufacturing yields are presented. 13 refs., 42 figs., 20 tabs

  7. Post test investigation of the single rod tests ESSI 1-11 on temperature escalation in PWR fuel rod simulators due to the Zircaloy/steam reaction

    International Nuclear Information System (INIS)

    This KfK-report describes the posttest investigation of the single rod tests ESSI-1 to ESSI-11. The objective of these tests was to investigate the temperature escalation behaviour of Zircaloy clad PWR-fuel rods in steam. The investigation of the temperature escalation is part of the program of out-of-pile experiments (CORA) performed within the frame work of the PNS Severe Fuel Damage Program. The experimental arrangement consisted of fuel rod simulator (central tungsten heater, UO2 ring pellets and Zircaloy cladding), Zircaloy shroud and fiber ceramic insulation. The introductory test ESSI-1 to ESSI-3 were scoping tests designed to obtain information on the temperature escalation of zircaloy in steam. ESSI-4 to ESSI-8 were run with increasing heating rates to investigate the influence of the oxide layer thickness at the start of the escalation. ESSI-9 to ESSI-11 were performed to investigate the influence of the insulation thickness on the escalation behaviour. In these tests we also learned that the gap between removed shroud and insulation has a remarkable influence due to heat removal by convection in the gap. After the test the fuel rod simulator was embedded into epoxy and cut by a diamond saw. The cross sections were photographed and investigated by metalograph microscope, SEM and EMP examinations. (orig./GL)

  8. Mathematical modelling of friction-vibration interactions of nuclear fuel rods

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2016-06-01

    Full Text Available Nuclear fuel rods (FRs are transverselly linked to each other by three spacer grid cells at several vertical levels inside a fuel assembly (FA. Vibration of FA components, caused by the motion of FA support plates in the reactor core, generates variable contact forces between FRs and spacer grid cells. Friction effects in contact surfaces have an influence on the expected lifetime period of nuclear FA in terms of FR cladding fretting wear. This paper introduces an original approach to mathematical modelling and simulation analysis of FR nonlinear vibrations and fretting wear taking into consideration friction forces at all levels of spacer grids.

  9. Determination of the Effectiveness of Control Rods in the VVER Reactor Fuel Assemblies

    International Nuclear Information System (INIS)

    The paper describes experiments done in homogeneous mock-ups of the fuel assemblies from the VVER Reactor (at one level of enrichment) to determine the effectiveness of absorbing systems comprising shim fuel assemblies or water cavities and of absorbing rods clad in jackets made of differing materials. The paper also gives data on some experiments that have been done in mock-ups of assemblies with differing levels of enrichment. These experiments make it possible to verify the methods used in calculation and to evaluate the prospects of using them for heterogeneous reactors. (author)

  10. Development of Application Technology of a Kagome Truss for a Fuel rod Support Structure

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ki Ju; Lee, Byung Chul; Kim, Pan Su [Chonnam National University, Gwangju (Korea, Republic of)

    2010-05-15

    The purpose of this work is to design a Wire-woven Bulk Kagome (WBK) cellular metal for a fuel rod support structure of a dual cooled fuel and to fabricate test samples. Design of WBK-based support - To analyze dynamic characteristics of a support structure with WBK core under side impact. - To specify strength of WBK to be used for the support. - To design strut length and diameter of WBK. Fabrication of the test samples - To assemble WBK samples from helically formed wires. - To braze WBK samples with side straps

  11. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  12. Model for the analysis of transitories and stability of a BWR reactor with fuel of thorium

    International Nuclear Information System (INIS)

    In this work it is described the thermo hydraulic and neutronic pattern used to simulate the behavior of a nucleus of thorium-uranium under different conditions of operation. The analysed nucleus was designed with base to assemblies that operate under the cover-seed concept. The pattern was proven to conditions of stationary state and transitory state. Here it is only presented the simulation of the one SCRAM manual and it is compared in the behavior of a nucleus with UO2. Additionally one carries out an analysis of stability taking into account the four corners that define the area of stability of the map flow-power and to conditions of 100% of flow and 100% of power. The module of stability is based on the pattern of Lahey and Podowsky to estimate the drops of pressure during a perturbation. It is concludes that the behavior of this nucleus is not very different to the one shown by the nuclei loaded with the fuel of UO2. (Author)

  13. Advanced and flexible genetic algorithms for BWR fuel loading pattern optimization

    International Nuclear Information System (INIS)

    This work proposes advances in the implementation of a flexible genetic algorithm (GA) for fuel loading pattern optimization for Boiling Water Reactors (BWRs). In order to avoid specific implementations of genetic operators and to obtain a more flexible treatment, a binary representation of the solution was implemented; this representation had to take into account that a little change in the genotype must correspond to a little change in the phenotype. An identifier number is assigned to each assembly by means of a Gray Code of 7 bits and the solution (the loading pattern) is represented by a binary chain of 777 bits of length. Another important contribution is the use of a Fitness Function which includes a Heuristic Function and an Objective Function. The Heuristic Function which is defined to give flexibility on the application of a set of positioning rules based on knowledge, and the Objective Function that contains all the parameters which qualify the neutronic and thermal hydraulic performances of each loading pattern. Experimental results illustrating the effectiveness and flexibility of this optimization algorithm are presented and discussed.

  14. Fuel design with low peak of local power for BWR reactors with increased nominal power; Diseno de un combustible con bajo pico de potencia local para reactores BWR con potencia nominal aumentada

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia C, R.; Montes, J.L.; Hernandez, J.L.; Ortiz, J.J.; Castillo, A. [ININ, 52750 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mrpc@nuclear.inin.mx

    2006-07-01

    The Federal Commission of Electricity recently announcement the beginning of the works related with the increase of the power to 120% of the original nominal one in the Boiling Water Reactors (BWR) of the Laguna Verde Central (CLV): In the National Institute of Nuclear Research (ININ) are carried out studies of the impact on the design of the recharge of derived fuel of this increase. One of the main effects of the power increase type that it is promoting, is the increment of the flow of generated vapor, what takes, to a bigger fraction of vacuum in the core presenting increased values of the maximum fraction to the limit, so much of the ratio of lineal heat generation (XFLPD) as of the ratio of critic power (MFLCPR). In the made studies, it is found that these fractions rise lineally with the increase of the nominal power. Considering that the reactors of the CLV at the moment operate to 105% of the original nominal power, it would imply an increment of the order of 13.35% in the XFLPD and in the MFLCPR operating to a nominal power of 120% of the original one. This would propitiate bigger problems to design appropriately the fuel cycle and the necessity, almost unavoidable, of to resort to a fuel assembly type more advanced for the recharges of the cores. As option, in the ININ the feasibility of continuing using the same type of it fuel assembles that one has come using recently in the CLV, the type GE12 is analyzed. To achieve it was outlined to diminish the peak factor of local power (LPPF) of the power cells that compose the fuel recharge in 13.35%. It was started of a fuel design previously used in the recharge of the unit 1 cycle 12 and it was re-design to use it in the recharge design of the cycle 13 of the unit 1, considering an increase to 120% of the original power and the same requirements of cycle extension. For the re-design of the fuel assembly cell it was used the PreDiCeldas computer program developed in the ININ. It was able to diminish the LPPF

  15. FREG-4: a computer program to predict the fuel-to-cladding heat transfer coefficient in accordance with the fuel-rod irradiation history

    International Nuclear Information System (INIS)

    Program FREG series calculate temperature distribution in a fuel rod and the stored energy based on the distribution. The temperature distribution is calculated accordance with the fuel rod irradiation history. The temperature in the fuel rod is severely influenced with gap heat transfer coefficients between fuel pellet surface and cladding inner surface. Enphasis is placed on how to find the gap heat transfer coefficients. FREG-4 is a version-up program of FREG-3. Major modification from FREG-3 is handlings of fission product gas release, which have influences on the gap heat transfer. FREG-4 distingushed fission-product isotopes remained in pellets and fission-product gaseous isotopes released from the pellets, and considers that the released isotopes are transported for plenums to balance whole fuel rod internal pressure and transformed into another isotopes due to decay and the nuetron absorptions. The present report describes modified models from FREG-3 and user's manual for FREG-4. (author)

  16. Transient non-boiling heat transfer in a fuel rod bundle during accidental power excursions

    International Nuclear Information System (INIS)

    The physical problem studied is the transient non-boiling heat transfer of a cylindrical fuel rod consisting of fuel, gap, and cladding to a steady, fully developed turbulent flow. The fuel pin is assumed to be located in the interior region of a subassembly with regular triangular or square arrangements. The turbulent velocity field as well as turbulent transport properties are specified as functions of the coordinates normal to the axial flow direction. The heat generation within the fuel may be specified as an arbitrary function of the three spatial coordinates and time. A digital computer program has been developed. On the basis of finite-difference techniques, to solve the governing partial differential equations with their associated subsidiary conditions. Results have been obtained for a series of exponential power transients of interest to safety of liquid-metal and water cooled nuclear reactors. The general physical features of transient convective heat transfer as explored by previous investigators have qualitatively been substantiated by the present analysis. Emphasis has been devoted to investigate the differences of heat-transfer (coefficient) results from multi-region analysis including a realistic fuel rod model and single-region analysis for the coolant region only. A comparison with the engineering relationships for turbulent liquid-metal cooling by Stein, which are an extension of the heat transfer coefficient concept to account for transient heat fluxes, clearly demonstrates that, at the parameters studied, Stein's approach tends to largely overestimate the convective heat transfer at early times

  17. Modeling of fuel-rod behavior during reactor power cycling and ramping experiments with computer code FRAPCON-2

    International Nuclear Information System (INIS)

    Modelling of fuel-rod behavior during reactor power cycling and ramping (including power-cooling mismatch experiments) with the computer code FRAPCON-2 is discussed. FRAPCON-2 computer calculations, using different mechanical models (Rigid Pellet, Deformable Pellet and Finite Element Mechanical Models) are compared with experimental results. The range of conditions over which FRAPCON-2 may be applied for PWR fuel rod behavior modelling during reactor power cycling and ramping are illustrated

  18. Process for identifying objects, particularly fuel rods of a nuclear reactor and device for carrying out the process

    International Nuclear Information System (INIS)

    In order to identify fuel rods of a nuclear reactor, ultrasonic seels are used, which consist of a polycrystalline material whose grain boundaries act as scatter centres when irradiated with ultrasonics and which cause reproducable scatter interference structures. By detecting and comparing the scatter interference structures of the high frequency ultrasonic backscatter signals, fuel rods which have been measured in the context of a zero picture of the reflectogram and provided with an ultrasonic seal can be identified again. (orig.)

  19. Spent Fuel Behaviour During Interim Storage

    International Nuclear Information System (INIS)

    Objective: Review of spent fuel data relevant for future storage in Spain Perform destructive and non-destructive examinations on irradiated and non-irradiated fuel rods relevant to Spanish spent fuel management. Research approach: Among the programmes initiated in the last years (finished or about to be finished) one may highlight the following ones: • Isotopic measurements on high burnup fuels: up to 75 GW·d·t(U)-1 PWR and 53 GW·d·t(U)-1 BWR peak values; • Mechanical tests on high burnup PWR (ZIRLO) cladding and BWR (Zry-2) cladding samples; • Mechanical tests on unirradiated ZIRLO rods. Influence of hydrides content; • Modelling of mechanical tests with unirradiated claddings; • Interim storage creep modelling; • Burnup measurement equipment; • Fuel database

  20. A methodology for the evaluation of fuel rod failures under transportation accidents

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, J.Y.R.; Machiels, A.J. [ANATECH, San Diego, CA (United States)]|[EPRI, Palo Alto (United States)

    2004-07-01

    Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

  1. A New Insight into Energy Distribution of Electrons in Fuel-Rod Gap in VVER-1000 Nuclear Reactor

    Science.gov (United States)

    Fereshteh, Golian; Ali, Pazirandeh; Saeed, Mohammadi

    2015-06-01

    In order to calculate the electron energy distribution in the fuel rod gap of a VVER-1000 nuclear reactor, the Fokker-Planck equation (FPE) governing the non-equilibrium behavior of electrons passing through the fuel-rod gap as an absorber has been solved in this paper. Besides, the Monte Carlo Geant4 code was employed to simulate the electron migration in the fuel-rod gap and the energy distribution of electrons was found. As for the results, the accuracy of the FPE was compared to the Geant4 code outcomes and a satisfactory agreement was found. Also, different percentage of the volatile and noble gas fission fragments produced in fission reactions in fuel rod, i.e. Krypton, Xenon, Iodine, Bromine, Rubidium and Cesium were employed so as to investigate their effects on the electrons' energy distribution. The present results show that most of the electrons in the fuel rod's gap were within the thermal energy limitation and the tail of the electron energy distribution was far from a Maxwellian distribution. The interesting outcome was that the electron energy distribution is slightly increased due to the accumulation of fission fragments in the gap. It should be noted that solving the FPE for the energy straggling electrons that are penetrating into the fuel-rod gap in the VVER-1000 nuclear reactor has been carried out for the first time using an analytical approach.

  2. Development of the ACP safeguards neutron counter for PWR spent fuel rods

    Science.gov (United States)

    Lee, Tae-Hoon; Menlove, Howard O.; Lee, Sang-Yoon; Kim, Ho-Dong

    2008-04-01

    An advanced neutron multiplicity counter has been developed for measuring spent fuel in the Advanced spent fuel Conditioning Process (ACP) at the Korea Atomic Energy Research Institute (KAERI). The counter uses passive neutron multiplicity counting to measure the 244Cm content in spent fuel. The input to the ACP process is spent fuel from pressurized water reactors (PWRs), and the high intensity of the gamma-ray exposure from spent fuel requires a careful design of the counter to measure the neutrons without gamma-ray interference. The nuclear safeguards for the ACP facility requires the measurement of the spent fuel input to the process and the Cm/Pu ratio for the plutonium mass accounting. This paper describes the first neutron counter that has been used to measure the neutron multiplicity distribution from spent fuel rods. Using multiple samples of PWR spent fuel rod-cuts, the singles (S), doubles (D), and triples (T) rates of the neutron distribution for the 244Cm nuclide were measured and calibration curves were produced. MCNPX code simulations were also performed to obtain the three counting rates and to compare them with the measurement results. The neutron source term was evaluated by using the ORIGEN-ARP code. The results showed systematic difference of 21-24% in the calibration graphs between the measured and simulation results. A possible source of the difference is that the burnup codes have a 244Cm uncertainty greater than ±15% and it would be systematic for all of the calibration samples. The S/D and D/T ratios are almost constant with an increment of the 244Cm mass, and this indicates that the bias is in the 244Cm neutron source calculation using the ORIGEN-ARP source code. The graphs of S/D and D/T ratios show excellent agreement between measurement and MCNPX simulation results.

  3. Development of Sirius facility that simulates void-reactivity feedback, and regional and core-wide stability estimation of natural circulation BWR

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, M.; Inada, F.; Yasuo, A. [Tokyo Electric Power Co., Inc., Central Research Institute (Japan)

    2001-07-01

    The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of fluid conditions, power distributions, and fuel rod thermal conductivity time constants, including the normal operating conditions of a typical natural circulation BWR. The results showed that there is a sufficiently wide stability margin under normal operating conditions, even when void-reactivity feedback is taken into account. (author)

  4. FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

    1981-01-01

    FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

  5. Description and characterization of BRPR series S-0, S-1, S-2, S-3, and S-4 demonstration fuel rods

    International Nuclear Information System (INIS)

    This report describes the process development, fabrication, and pre-irradiation characterization of the demonstration fuel rods for irradiation in the Big Rock Point Reactor as part of the DOE-sponsored Fuel Performance Improvement Program (FPIP). The fuel rods represent advanced designs that are expected to exhibit improved performance with respect to pellet-cladding-interaction and the attainment of extended burnup. Whereas other design variations are described, the primary fuel concepts being evaluated as part of the FPIP are an annular-coated-pressurized design and, at a more modest level, a sphere-pac design. A solid-pellet reference design provides the basis for comparing irradiation behavior

  6. Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction: Tests ESSI-1,2,3

    International Nuclear Information System (INIS)

    This report discusses the test conduct, results, and posttest appearance of three scoping tests (ESSI-1,2,3) investigating temperature escalation in zircaloy clad fuel rods. The experiments are part of an out-of-pile program using electrically heated fuel rod simulators to investigate PWR fuel element behavior up to temperatures of 20000C. These experiments are part of the PNS Severe Fuel Damage Program. The temperature escalation is caused by the exothermal zircaloy/steam reaction, whose reaction rate increases exponentially with the temperature. The tests were performed using different initial oxide layers as a major parameter, obtained by varying the heatup rates and steam exposure times. (orig./RW)

  7. Assessing the Effect of Fuel Burnup on Control Rod Worth for HEU and LEU Cores of Gharr-1

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

    Full Text Available An important parameter in the design and analysis of a nuclear reactor is the reactivity worth of the control rod which is a measure of the efficiency of the control rod to absorb excess reactivity. During reactor operation, the control rod worth is affected by factors such as the fuel burnup, Xenon concentration, Samarium concentration and the position of the control rod in the core. This study investigates the effect of fuel burnup on the control rod worth by comparing results of a fresh and an irradiated core of Ghana's Miniature Neutron Source Reactor for both HEU and LEU cores. In this study, two codes have been utilized namely BURNPRO for fuel burnup calculation and MCNP5 which uses densities of actinides of the irradiated fuel obtained from BURNPRO. Results showed a decrease of the control rod worth with burnup for the LEU while rod worth increased with burnup for the HEU core. The average thermal flux in both inner and outer irradiation sites also decreased significantly with burnup for both cores.

  8. Serpent: an alternative for the nuclear fuel cells analysis of a BWR; SERPENT: una alternativa para el analisis de celdas de combustible nuclear de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silva A, L.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: lidi.s.albarran@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In the last ten years the diverse research groups in nuclear engineering of the Universidad Nacional Autonoma de Mexico and Instituto Politecnico Nacional (UNAM, IPN), as of research (Instituto Nacional de Investigaciones Nucleares, ININ) as well as the personnel of the Nuclear Plant Management of the Comision Federal de Electricidad have been using the codes Helios and /or CASMO-4 in the generation of cross sections (X S) of nuclear fuel cells of the cores corresponding to the Units 1 and 2 of the nuclear power plant of Laguna Verde. Both codes belong to the Studsvik-Scandpower Company who receives the payment for the use and their respective maintenance. In recent years, the code Serpent appears among the nuclear community distributed by the OECD/Nea which does not has cost neither in its use neither in its maintenance. The code is based on the Monte Carlo method and makes use of the processing in parallel. In the Escuela Superior de Fisica y Matematicas of the IPN, the personnel has accumulated certain experience in the use of Serpent under the direction of personal of the ININ; of this experience have been obtained for diverse fuel burned, the infinite multiplication factor for three cells of nuclear fuel, without control bar and with control bar for a known thermodynamic state fixed by: a) the fuel temperature (T{sub f}), b) the moderator temperature (T{sub m}) and c) the vacuums fraction (α). Although was not realized any comparison with the X S that the codes Helios and CASMO-4 generate, the results obtained for the infinite multiplication factor show the prospective tendencies with regard to the fuel burned so much in the case in that is not present the control bar like when it is. The results are encouraging and motivate to the study group to continue with the X S generation of a core in order to build the respective library of nuclear data as a following step and this can be used for the codes PARCS, of USA NRC, DYN3D of HZDR, or others developed locally

  9. Proceedings of the specialist meeting on nuclear fuel and control rods: operating experience, design evolution and safety aspects

    International Nuclear Information System (INIS)

    Design and management of nuclear fuel has undergone a strong evolution process during past years. The increase of the operating cycle length and of the discharge burnup has led to the use of more advanced fuel designs, as well as to the adoption of fuel efficient operational strategies. The analysis of recent operational experience highlighted a number of issues related to nuclear fuel and control rod events raising concerns about the safety aspects of these new designs and operational strategies, which led to the organisation of this Specialists Meeting on fuel and control rod issues. The meeting was intended to provide a forum for the exchange of information on lessons learned and safety concern related to operating experience with fuel and control rods (degradation, reliability, experience with high burnup fuel, and others). After an opening session 6 papers), this meeting was subdivided into four sessions: Operating experience and safety concern (technical session I - 6 papers), Fuel performance and operational issues (technical session II - 7 papers), Control rod issues (technical session III - 9 papers), Improvement of fuel design (technical session IV.A - 4 papers), Improvement on fuel fabrication and core management (technical session IV.B - 6 papers)

  10. Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors

    International Nuclear Information System (INIS)

    In the National Institute of Nuclear Research (ININ) a methodology is developed to optimize the design of cells 10x10 of assemble fuels for reactors of water in boil or BWR. It was proposed a lineal calculation formula based on a coefficients matrix (of the change reason of the relative power due to changes in the enrichment of U-235) for estimate the relative powers by pin of a cell. With this it was developed the computer program of fast calculation named PreDiCeldas. The one which by means of a simple search algorithm allows to minimize the relative power peak maximum of cell or LPPF. This is achieved varying the distribution of U-235 inside the cell, maintaining in turn fixed its average enrichment. The accuracy in the estimation of the relative powers for pin is of the order from 1.9% when comparing it with results of the 'best estimate' HELIOS code. With the PreDiCeldas it was possible, at one minimum time of calculation, to re-design a reference cell diminishing the LPPF, to the beginning of the life, of 1.44 to a value of 1.31. With the cell design with low LPPF is sought to even design cycles but extensive that those reached at the moment in the BWR of the Laguna Verde Central. (Author)

  11. Experimental Fission Gas Release Determination at High Burnup by Means of Gamma Measurements on Fuel Rods in OL2

    International Nuclear Information System (INIS)

    This article presents the results from the gamma measurements performed on a selection of fuel rods from two SVEA-96 Optima fuel assemblies in Olkiluoto unit 2 (OL2) during February 2008. The measurements were funded by Teollisuuden Voima Oyj (TVO) and carried out by Westinghouse Electric Sweden AB (WSE). The goal of the measurements was to obtain plant specific fission gas release data from OL2 which will later be used to support TVO's burnup increase. The measurements were performed by means of detecting and recording information on gamma rays emanating from radioactive fission products within the fuel rods. A fuel assembly under operation in the reactor will be subject to fission gas release, meaning that gaseous fission products in the fuel matrix will leak out into the fuel rod free volumes, including the upper fuel rod plenum. The magnitude of fission gas release is closely related to the operation conditions, such as burnup and the pellet power/temperature history. Fission gas which is released into the plenum leads to an increased pressure within the fuel rod. Since the gas plenum is a volume free of fuel pellets, radioactive gases present here are therefore relatively easy to measure. A suitable gaseous radioactive fission product is 85Kr which has a gamma energy of 514 keV, and a half life of 10.8 years. Measuring the amount of 85Kr can then be used to quantify the amount of fission gases release during operation. Other sources of gamma rays in the plenum with similar energies are the β+-emitting 58Co and the 106Ru. The half lives of 58Co and 106Ru are 71 days and 372 days, respectively, with corresponding gamma ray energies of 511 keV and 512 keV, respectively. In total, three different gamma rays with similar energies must be resolved by the detector system. In order to perform the measurements, 58Co must have decayed to an extent that allows the 514 keV line to be resolved from the 511 keV positron annihilation gamma ray. The 106Ru is, on the other

  12. Study of transient heat transfer in a fuel rod 3D, in a situation of unplanned shutdown of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Martins, Rodolfo Ienny; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: rodolfoienny@gmail.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The study, in situations involving accidents, of heat transfer in fuel rods is of known importance, since it can be used to predict the temperature limits in designing a nuclear reactor, to assist in making more efficient fuel rods, and to increase the knowledge about the behavior of the reactor's components, a crucial aspect for safety analysis. This study was conducted using as parameter the fuel rod that has the highest average power in a typical PWR reactor. For this, we developed a program (Fuel{sub R}od{sub 3}D) in Fortran language using the Finite Elements Method (FEM) for the discretization of a fuel rod and coolant channel, in order to study the temperature distribution in both the fuel rod and the coolant channel. Transient parameters were coupled to the heat transfer equations in order to obtain details of the behavior of the rod and the channel, which allows the analysis of the temperature distribution and its change over time. This work aims to present a study case of an accident where there is a lack of energy in the reactor's coolant pumps and in the diesel engines, resulting in an unplanned shutdown of the reactor. In order to achieve the intended goal, the present work was divided as follows: a short introduction about heat transfer, including the equations concerning the fuel rod and the energy equation in the channel, an explanation about how the verification of the Fuel{sub R}od{sub 3}D program was made, and the analysis of the results. (author)

  13. Simulation of accident and normal fuel rod work with Zr-cladding

    International Nuclear Information System (INIS)

    The technique of simulation of heat-physics, strength and safety characteristics of reactor RBMK and WWER rods under steady-state, transient and accident conditions is presented. That technique is used in mechanic and heat physics codes PULSAR-2 and STALACTITE. Simulation in both full scale and the most stress-loading part of cladding statement under accident conditions are considered. In this zone local swelling and cladding failure are possible. The accident simulation is based on the mechanical creep-plasticity problem solution in three-dimensional approach. The local cladding swelling is initiated with determining of little hot spot on the clad with several degrees temperature departure from average value. Mechanical problem is solved by finite elements method. Interaction of Zr with steam is taken in to account. Fuel and cladding melting, shortness and dispersion formation processes are simulated under subsequent rods warming up. (author). 2 refs., 6 figs

  14. A 2D-3D FEM approach of fuel rod thermomechanical behaviour during a RIA

    International Nuclear Information System (INIS)

    For better understanding of the fuel rod behaviour during a RIA and to extrapolate the CABRI tests results to PWR conditions, a pellet and its corresponding cladding part have been modelled by means of a 2D axisymmetric meshing, with EDF's finite element code ASTER. The pellet rim region, which is modelled with a 3D meshing, is represented in the global 2D-model with an equivalent homogenized material. The stress distribution is calculated by applying a thermal radial profile computed with the CEA/IPSN SCANAIR code. Then, the local stresses are determined in the rim region, in the neighbourhood of a gas bubble. This 2D-3D FEM approach has been applied successively to REP Na1 rod, at the time and location of the first failure, and to the postulated RCCA ejection accident in a PWR. (R.P.)

  15. A finite element method with contact for tensile analysis in fuel rods

    International Nuclear Information System (INIS)

    Elements for mechanical analysis of fuel rod of a PWR type reactor, are presented. The rod, consists basically in a cylindrical coating of zircalloy which contains pilling of UO2 pellets, is submitted to strong internal and external pressures, intense temperature gradients and neutron flux. These conditions lead several phenomena in the pellet (swelling, fracture, densification, creep) and in the cladding (embrittlement, corrosion, creep) which undergo deformations leading them to contact the restriction for the interpenetration is included in the problem without restriction by Lagrange multipliers. Considering a non-linear problem, due to the surface of contact to be not known a priori, the numerical solutions were obtained using the finite element method. (M.C.K.)

  16. Supplemental description of ROSA-IV/LSTF with No.1 simulated fuel-rod assembly

    International Nuclear Information System (INIS)

    Forty-two integral simulation tests of PWR small break LOCA (loss-of-coolant accident) and transient were conducted at the ROSA-IV Large-Scale Test Facility (LSTF) with the No.1 simulated fuel-rod assembly between March 1985 and August 1988. Described in the report are supplemental information on modifications of the system hardware and measuring systems, results of system characteristics tests including the initial fluid mass inventory and heat loss distribution for the primary system, and thermal properties for the heater rod materials. These are necessary to establish the correct boundary conditions of each LSTF experiment with the No.1 core assembly in addition to the system data given in the system description report (JAERI-M 84-237). (author)

  17. Measurement of Fresh Fuel Rods to Demonstrate Compliance with Criticality Safety Limits

    Energy Technology Data Exchange (ETDEWEB)

    Miko, David K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-03

    In order to operate TA-66 as a radiological facility with the quantity of nuclear material required to fulfil its mission, a criticality safety evaluation was required. This evaluation defined the control parameters for operations at the facility. The resulting evaluation for TA-66 placed limits on the amount of SNM, as well as other materials such as beryllium. In addition, there is a limit on the number of uranium fuel rods allowed subject to enrichment, outer diameter, and overall length restrictions. The enrichments for the rods to be shipped to TA-66 were documented in LA-UR-13-23581, but the outer diameter and length were not documented. This report provides this information.

  18. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    Energy Technology Data Exchange (ETDEWEB)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR.

  19. PIE of the second fuel rod from the LOCA experiment (IFA-650.2)

    Energy Technology Data Exchange (ETDEWEB)

    Oberlaender, B.C.; Jenssen, H.K.; Espeland, M.; Solum, N.O.

    2005-07-01

    The LOCA experiment on the second rod (IFA-650.2) a fresh, low-tin Zr-4, pressurised PWR rod was carried out in May 2004. The main objective was to produce ballooning, to determine the time to burst and to assess the material oxidation and hydriding kinetics. The rod pressure at hot conditions and peak PCT were 70 bar and 1050 C, respectively. To document the effect of the LOCA testing on the cladding, rod 2 was subjected in PIE to visual inspection, profilometry and metallography. On visual inspection the clad shows a typical balloon. In the region of max ballooning the clad shows a 35 mm long, < 20 mm burst opening. In the balloon region the outer clad diameter increased by <35% and locally the wall thickness reduction is >55%. The entire rod is covered with a black oxide layer. Below and above the split opening the continuous oxide layer was 40 to 50mum both on water and fuel side of the clad. At the locations of the upper and lower cladding thermocouples the oxide thickness was in the range 24-27 mum. Widmanstaetten structure is seen in the bulk of the clad and confirms the high temperature experienced. A some 40mum wide, hard and brittle zone with oxygen rich globular alpha-grains is found both at the outer and the inner edge of the clad in the balloon region. The zone is widest near the axial split (crack). In the clad few, arbitrary oriented hydride platelets are observed in the balloon area. (Author)

  20. A High Fidelity Multiphysics Framework for Modeling CRUD Deposition on PWR Fuel Rods

    Science.gov (United States)

    Walter, Daniel John

    Corrosion products on the fuel cladding surfaces within pressurized water reactor fuel assemblies have had a significant impact on reactor operation. These types of deposits are referred to as CRUD and can lead to power shifts, as a consequence of the accumulation of solid boron phases on the fuel rod surfaces. Corrosion deposits can also lead to fuel failure resulting from localized corrosion, where the increased thermal resistance of the deposit leads to higher cladding temperatures. The prediction of these occurrences requires a comprehensive model of local thermal hydraulic and chemical processes occurring in close proximity to the cladding surface, as well as their driving factors. Such factors include the rod power distribution, coolant corrosion product concentration, as well as the feedbacks between heat transfer, fluid dynamics, chemistry, and neutronics. To correctly capture the coupled physics and corresponding feedbacks, a high fidelity framework is developed that predicts three-dimensional CRUD deposition on a rod-by-rod basis. Multiphysics boundary conditions resulting from the coupling of heat transfer, fluid dynamics, coolant chemistry, CRUD deposition, neutron transport, and nuclide transmutation inform the CRUD deposition solver. Through systematic parametric sensitivity studies of the CRUD property inputs, coupled boundary conditions, and multiphysics feedback mechanisms, the most important variables of multiphysics CRUD modeling are identified. Moreover, the modeling framework is challenged with a blind comparison of plant data to predictions by a simulation of a sub-assembly within the Seabrook nuclear plant that experienced CRUD induced fuel failures. The physics within the computational framework are loosely coupled via an operator-splitting technique. A control theory approach is adopted to determine the temporal discretization at which to execute a data transfer from one physics to another. The coupled stepsize selection is viewed as a

  1. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

  2. Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

    International Nuclear Information System (INIS)

    The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO2 and PuO2-UO2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO2 rods at two enrichments (2.35 wt % and 4.31 wt % 235U) and on mixed fuel-water assemblies of UO2 and PuO2-UO2 rods containing 4.31 wt % 235U and 2 wt % PuO2 in natural UO2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

  3. A cone beam computed tomography inspection method for fuel rod cladding tubes

    International Nuclear Information System (INIS)

    Fuel rods in nuclear power plants consist of UO2 pellets enclosed in Zirconium alloy (Zircaloy) cladding tube, which is composed of a body and a plug. The body is manufactured separately from the plug and, before its use, the plug is welded with the body. It is vitally important for the welding zone to remain free from defects after the fuel pellets are loaded into the cladding tube to prevent the radioactive fission products from leaking. X-ray computed tomography (CT) is in principle a feasible inspection method for the welding zone, but it faces several challenges due to the high attenuation of Zircaloy. In this paper, a cone beam CT method is proposed to address these issues and perform the welding flaw inspection. A Zircaloy compensator is adopted to narrow the signal range, a structure-based background removal technique to reveal the defects, a linear extension technique to determine the reference X-ray intensity signal and FDK algorithm to reconstruct the slice images. A prototype system, based on X-ray tube source and flat panel detector, has been developed and the experiments in this system have demonstrated that the welding void and the incomplete joint penetrations could be detected by this method. This approach may find applications in the quality control of nuclear fuel rods.

  4. End Plug Welding of New FM Cladding Tube for SFR Metallic Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Ferritic/martensitic (FM) steel such as HT9 has been selected for a SFR fuel cladding tube material because of the excellent irradiation characteristics. But the HT9 material is not enough to satisfy the discharge burnup goal due to the high coolant outlet temperature and the low creep resistance. Recently new FM steel is under development to improve HT9 material properties. At KAERI, development of new FM steels, named FC92B and FC92N, has been in progress to improve thermal creep resistance. In this study, the qualification test of the end plug welding of new FM steels, FC92B and FC92N, for SFR metallic fuel rods was carried out based on the developed HT9 welding technology. Through the qualification test, the end plug welding of new FM steels was well verified. Through the qualification test, the weld quality of end plug welding of new FM cladding tubes, FC92B and FC92N, was evaluated according to the qualification test plan and satisfied the requirements on the weld. Consequently, the qualified welding process is ready to produce SFR metallic fuel rods.

  5. A cone beam computed tomography inspection method for fuel rod cladding tubes

    Science.gov (United States)

    Fu, Jian; Tan, Renbo; Wang, Qianli; Deng, Jingshan; Liu, Ming

    2012-10-01

    Fuel rods in nuclear power plants consist of UO2 pellets enclosed in Zirconium alloy (Zircaloy) cladding tube, which is composed of a body and a plug. The body is manufactured separately from the plug and, before its use, the plug is welded with the body. It is vitally important for the welding zone to remain free from defects after the fuel pellets are loaded into the cladding tube to prevent the radioactive fission products from leaking. X-ray computed tomography (CT) is in principle a feasible inspection method for the welding zone, but it faces several challenges due to the high attenuation of Zircaloy. In this paper, a cone beam CT method is proposed to address these issues and perform the welding flaw inspection. A Zircaloy compensator is adopted to narrow the signal range, a structure-based background removal technique to reveal the defects, a linear extension technique to determine the reference X-ray intensity signal and FDK algorithm to reconstruct the slice images. A prototype system, based on X-ray tube source and flat panel detector, has been developed and the experiments in this system have demonstrated that the welding void and the incomplete joint penetrations could be detected by this method. This approach may find applications in the quality control of nuclear fuel rods.

  6. Model for estimation of the amount of dispersed fuel during operation of BWRs with defected rods

    International Nuclear Information System (INIS)

    Two different models for estimations of the amount of tramp uranium in the cores in BWRs are presented. One of the models is useful to estimate the amount of tramp uranium during operation with non-defected fuel. The measured release rates of Xe-133 and I-131 are compared with those calculated, by the ORIGEN2-code. A nomogram has been designed for easy use of the model. The second model is useful to estimate the accumulation rate of dispersed fuel on the core surfaces during operation with defected and degraded fuel rods. The model is empirical, the measured Np-239 concentration in the reactor water is used as indicator on dispersed uranium dioxide. The model has shown to be reliable and useful for consequence analyses of recorded failures. The results of the estimations with this model can be compared with the results of the first model after the defected rods have been exchanged. Examples on the use of the models are given and discussed. (authors). 4 figs., 1 tab., 4 refs

  7. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system

  8. A statistical approach to pellet stack movement during fuel rod ballooning

    International Nuclear Information System (INIS)

    The principal factor which determines the burst strain of a ballooning PWR fuel rod during a postulated loss-of-coolant accident is the magnitude of the circumferential temperature variation around the Zircaloy cladding. It is currently accepted that heat flux asymmetries associated with pellet eccentricity within the ballooning clad are the main source of such temperature variations. This paper presents an analysis of fuel pellet stack configurational changes which may accompany clad ballooning and consequently influence the development of clad azimuthal temperature gradients. The basis of the model rests on the premise that a stack of fuel pellets whose end-faces are not orthogonal to the cylinder axis but which are constrained within the cladding will develop inter-pellet gaps. During ballooning, closure of the gaps results in a lateral movement of the stack and provides a mechanism by which the central pellets are able to maintain unit eccentricity until the clad has ballooned to a diametral strain at which the inter-pellet gaps are closed. To describe the effect on clad deformation of the pellet-clad configuration changes, a statistical approach to stack movement has been coupled to the TAPSWEL code. It is shown that the distribution of burst strains in the MT-3 rods from the in-pile LOCA simulation test can be predicted using geometric parameters measured on archive pellets. The observed variation of the burst strains exhibited by the MT-3 rods is shown to be reasonably consistent with the model prediction based on the pellet stack displacement concept. (author)

  9. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

  10. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation

  11. Fuel rod and core materials investigations related to LWR extended burnup operation

    Science.gov (United States)

    Kolstad, Erik; Vitanza, Carlo

    1992-06-01

    The paper deals with tests and recent measurements related to extended burnup fuel performance and describes test facilities and results in the areas of waterside cladding corrosion and irradiation-assisted stress corrosion cracking (IASCC). Fuel temperature data suggest a gradual degradation of UO 2 thermal conductivity with exposure in the range 6-8% per 10 MWd/kgUO 2 at temperatures below 700°C. The effect on the fuel microstructure of interlinkage and resintering phenomena is shown by measuring the surface-to-volume ( S/ V) ratio of the fuel. Changes in S/V with burnup are correlated to power rating and fuel operating temperature. No evidence was found of enhanced fission gas release during load-follow operation in the burnup range 25-45 MWd/kgUO 2. The effect of high lithium concentration (high pH) on the corrosion behaviour of pre-irradiated high burnup Zircaloy-4 fuel rods subjected either to nucleate boiling or to one-phase cooling conditions was studied. The oxide thickness growth rates measured at an average burnup up to 40 MWd/kgUO 2 are consistent with literature data and show no evidence of corrosion enhancement due to the high lithium content and little effect of cooling regime. A test facility for exploring the effects of environmental variables on IASCC behaviour of in-core structural materials is described.

  12. A fast and compact Fuel Rod Performance Simulator code for predictive, interpretive and educational purpose

    International Nuclear Information System (INIS)

    A new Fuel rod Performance Simulator code FRPS has been developed, tested and benchmarked and is now available in different versions. The user may choose between the batch version INTERPIN producing results in form of listings or beforehand defined plots, or the interactive simulator code SIMSIM which is stepping through a power history under the control of user. Both versions are presently running on minicomputers and PC:s using EGA-Graphics. A third version is the implementation in a Studsvik Compact Simulator with FRPS being one of its various modules receiving the dynamic inputs from the simulator

  13. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    International Nuclear Information System (INIS)

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  14. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D.L.; Reymann, G.A. (comps.)

    1979-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

  15. Matpro--version 10: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Reymann, G.A. (comp.)

    1978-02-01

    The materials properties correlations and computer subcodes (MATPRO--Version 10) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory are described. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

  16. Mechanical analysis of the fuel rod with the KWU code SENATOR

    International Nuclear Information System (INIS)

    For PCMI analyses of power ramps it is necessary to have good theoretical information concerning the mechanical condition of the clad - especially the hoop stress - during and after a power ramp. As a tool for such analyses KWU has developed the computer code SENATOR which covers the mechanical analysis in a slice of a fuel rod. For SENATOR calculations the initial conditions of a power ramp are evaluated by the fuel rod code CARO. The first paper about the SENATOR code was presented at the 6th SMIRT-Conference (1981). In the meantime the code has been refined and an improved version is described in this paper. The mechanical models in the code can be characterized by the following features: The pellet is divided into three concentric zones with different axial boundary conditions. A sub-model defines the axial coupling between fuel and clad. The SENATOR code calculates average stresses and strains in the clad. Combined with concentration factors - as derived from FEM-analyses - local stresses and strains can be calculated. In the improved version an anisotropic Young's modulus of the fuel has been introduced, which is considerably reduced compared to the value of the bulk material. This modification can be explained by the mechanical behaviour of the cracked pellet. In the fuel creep law another term has been added, which is due to local creep effects during closure of the pellet cracks especially at the beginning of relaxation. The SENATOR code has been calibrated against 8 Halden ramp tests, which cover the range of centerline temperature up to 2200 deg. C, the burnup range up to 20 MWd/kg(U), and the range in gap size up to 200 μm. In the Halden experiments the axial clad elongation is measured during and after the power ramp and for the SENATOR calibration the predicted axial elongations are fitted to the measurements. (author)

  17. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO2-PuO2 fuel

    International Nuclear Information System (INIS)

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs

  18. Method for length reduction of at least one irradiated fuel rod and cladding required to carry out the method

    International Nuclear Information System (INIS)

    For the length reduction of irradiated fuel rods, the conventional method is to first put the fuel rod into a straight sealing tube and then to reduce the length by deforming it, i.e. the sealing tube and the fuel rod contained are coiled up. This method for length reduction, however, creates problems with the production of storage containers that are to take up the wound piece. To avoid such difficulties, it has been suggested to have with each sealing tube at least one deformation element that covers part of the fuel rod of a certain length like a sleeve and that is connected gastight to the sealing tube; it remaind gastight during deformation. The fuel rod is bent in the deformation area determined by the deformation element at an angle of 1800, the deformation element being deformed according to the buckling, and the sealing tube remains for most of its length rectilinear. The deformation element may be separately modeled or a one-piece component of the sealing tube. In the latter case the cross sections of the deformation element and of the sealing tube can be the same. (orig.)

  19. Temperature distribution on fuel rods: a study on the effect of eccentricity in the position of UO2 pellets

    International Nuclear Information System (INIS)

    This work proposes the development of a method of solving equations of heat transfer applied in fuel rods using the finite element method, in order to evaluate the performance and safety of the nuclear system. Was prepared in a Fortran program to evaluate the equations governing the problem, the boundary conditions and apply the properties of materials on a steady state. This program uses the mesh generation input and graphical output generated by the program GID. The method was validated against the analytical solution found in the book Todreas and Kazimi with error less than 0.2% and with respect to the improved analytical solution of Nijsing for axisymmetry rod and eccentricity rod with error less than a 3.6%. Applications have been developed with the use of correlations for properties with the temperature dependence of resolution axisymmetry rod and the resolution of a rod with eccentricity. The method developed, should it be implemented, would allow the assessment of fuel rods in the given situations and other scenarios, as well as adding a tool of substantial value in the analysis of rods. (author)

  20. Synthesis and Analysis of Alpha Silicon Carbide Components for Encapsulation of Fuel Rods and Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Kevin M. McHugh; John E. Garnier; George W. Griffith

    2011-09-01

    The chemical, mechanical and thermal properties of silicon carbide (SiC) along with its low neutron activation and stability in a radiation field make it an attractive material for encapsulating fuel rods and fuel pellets. The alpha phase (6H) is particularly stable. Unfortunately, it requires very high temperature processing and is not readily available in fibers or near-net shapes. This paper describes an investigation to fabricate a-SiC as thin films, fibers and near-net-shape products by direct conversion of carbon using silicon monoxide vapor at temperatures less than 1700 C. In addition, experiments to nucleate the alpha phase during pyrolysis of polysilazane, are also described. Structure and composition were characterized using scanning electron microscopy, energy dispersive spectroscopy and X-ray diffraction. Preliminary tensile property analysis of fibers was also performed.

  1. The State of the Art of the Decladding Method for the Spent Nuclear Fuel Rods

    International Nuclear Information System (INIS)

    Our country's energy consumption is increasing day after day even though it relies on imports for more than 95 percent of its energy needs. In this circumstances, the atomic energy is a promising alternative to solve the problem of an energy security and an environmental preservation simultaneously. However, nuclear power produces spent fuel which is a highly radioactive waste. For a reliable and effective management of the spent fuel, the ACP(Advanced Spent Fuel Conditioning Process) is being developed at the KAERI. As a first state equipment of the ACP, a decladding machine is used to separate spent fuel rod into the UO2 pellets and hulls. This technical report aims to analyze existing decladding methods, and then, find a suitable decladding mechanism for the ACP. Many studies on the decladding of spent fuel can be categorized two approaches: chemical approach and mechanical one. In this report, we concentrated on the mechanical decladding approach. We developed engineering scale decladding device(20 kgHM/batch) and evaluated the performance through the verification experiments. We expect that this technical report helps in developing a scale-up equipment and technology

  2. Welding process optimization of nuclear fuel rod using TIG technique for fuel irradiation tests

    International Nuclear Information System (INIS)

    The irradiation test of fuel elements was planned for the evaluation of nuclear fuel performance. To establish fabrication process satisfying the requirements of irradiation test, orbital-TIG welding system for fuel elements was developed, and preliminary welding experiments for optimizing process conditions of fuel element was performed. Fuel elements with 9.5mm diameter and 0.6mm wall thickness of cladding tubes and end caps have been used and optimum conditions of endcap welding have been selected. In this study, the qualification test was performed by tensile tests, helium leak inspections, and metallography examinations to qualify the end cap welding procedure. The soundness of the welds quality of nuclear fuel elements has been confirmed by mechanical tests and microstructural examinations

  3. BOR-60 reactor as an instrument for experimental substantiation of fuel rods for advanced NPPs

    International Nuclear Information System (INIS)

    Full text: The BOR-60 fast test reactor is actually the only facility of this type in the world that has been in reliable and continuous operation for about 35 years. One of the principle reactor tasks is irradiation of advanced fuel and structural materials in different conditions. Inside the reactor the materials can be irradiated in any core and reflector cell except seven cells used for control rods. The number of fuel assemblies loaded into the reactor can vary from 85 to 124 depending on the burnup, core configuration and fuel properties. Due to the reactor design, the core dimensions can be widely changed allowing accommodation of no less than 20 experimental assemblies in different reactor cells. The neutron flux value in individual cells can vary more than 3 times at the maximum value of 3.7·1015 n/cm2s. Thus various fuel compositions can be loaded into the reactor and practically any burnups can achieve. Based on the long-term investigation of the reactor characteristics, we studied the reactor behavior in different conditions, developed a set of the verified codes and different procedures for the on-line reactor maintenance and performance of the wide scope of experiments. A set of specialized testing facilities consisting of capsule units and dismountable assemblies are used for irradiation of the wide range of materials and items at different conditions. The advantages of these facilities are their simplicity and possibility of installation in any core and reflector cell. In addition to the precision calculations of the irradiation conditions there is also a possibility for monitoring the neutron flux and temperature. A special thermometric channel available in the core allows accommodation of the experimental facilities and output of information of the irradiation conditions by 30-50 communication lines. It was required to develop a series of independent instrumented capsule-loops, special instrumented fuel assemblies etc. to be used in the channel

  4. Design of the Testing Set-up for a Nuclear Fuel Rod by Neutron Radiography at CARR

    Science.gov (United States)

    Wei, Guohai; Han, Songbai; Wang, Hongli; Hao, Lijie; Wu, Meimei; He, Linfeng; Wang, Yu; Liu, Yuntao; Sun, Kai; Chen, Dongfeng

    In this paper, an experimental set-up dedicated to non-destructively test a 15cm-long Pressurized Water Reactor (PWR) nuclear fuel rod by neutron radiography (NR) is described. It consists of three parts: transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo Simulation by the MCNP code. The material for the shell of the transport container was chosen to be lead with the thickness of 13 cm. Also, the mechanical devices were designed to control fuel rod movement inside the container. The imaging block was designed as the exposure platform, with three openings for the neutron beam, neutron converter foil, and specimen. Development and application of this experimental set-up will help gain much experience for investigating the actual irradiated fuel rod by neutron radiography at CARR in the future.

  5. Post-irradiation examination of Al-61 wt% U3Si fuel rods from the NRU reactor

    International Nuclear Information System (INIS)

    This paper describes the post-irradiation examination of 4 intact low enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D2O coolant inlet temperature 37E C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 : m thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 : m thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on Al-61 wt% U3Si fuel irradiated in the NRU reactor. (author)

  6. Development of Fuel Rod Puncturing and Fission Gas Collection System for Post Irradiation Examination in Hot cell

    International Nuclear Information System (INIS)

    After a fuel pin and small fuel rig for R/D were irradiated in a research reactor, a small amount of the fission gas exists into the internal void in a pin and a rig. A steel needle was not useful for these small fuel rods. Alternatively, the laser puncturing technique was developed to solve the measurement of a small amount of fission gas. This system is considered very rare equipment in other countries. A fine pressure gauge and strong vacuum system were installed, and the chamber volume was reduced as small as possible. A fiber laser was used for easy operation. To measure the small fission gas inventory in a fuel rod, a laser puncturing method was introduced. The reliability of the pressure measurement was found to have an error of 1% after a test of four sample bottles with different pressures. In the test of six sample rods, an under estimation was found because of a different pressure gauge was used in the rod preparation. However, there was consistency between the results of some rods. We need to check the properties and mechanical differences of each gauge

  7. Analytical investigation of the properties of the neutron noise induced by vibrating absorber and fuel rods

    International Nuclear Information System (INIS)

    Analytical solution methods for the neutron noise in a one-dimensional multi-region system in two-group theory, which have so far been based on the adjoint function technique, are extended here to using the forward Green's function technique. The forward Green's functions were calculated analytically for a noise source in a core surrounded by reflector regions at both sides. It is shown that with symbolic computation methods, the forward Green's function can be used for the calculation of the space- and frequency-dependent noise in the first order approximation for arbitrary noise sources which have an analytical representation. The properties of the induced neutron noise were investigated for vibrations of both absorbers and fuel assemblies, with two representations of the noise sources: a point-like source which corresponds to the vibrations of a fuel rod, and a finite width source which corresponds to vibrations of a fuel assembly. The contributions of the components induced by the fluctuations of the various types of macroscopic cross sections in the total noise are also discussed and the information content of the noise in the fast group is explored for the identification of fuel assembly vibrations. (orig.)

  8. Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

    Directory of Open Access Journals (Sweden)

    Amir Zacarias Mesquita

    2013-01-01

    Full Text Available In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF. At this time, the small increase in heat flux or in the refrigerant inlet temperature at the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout. This paper describes the experiments conducted to detect critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN. It is concluded that the use of displacement transducer is the most efficient technique for detecting critical heat flux in nuclear simulators heated by electric current in open pool.

  9. Elastic analysis of thermal gradient bowing in rod-type fuel elements subjected to axial thrust (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Newman, J.B.

    1968-01-01

    Thermal radient bowing of rod type fuel elements can be analyzed in terms of the deflections of a precurved beam. The fundamental aspects of an analysis of axially compressed multispan beams are given. Elasticity of supports in both axial and transverse directions is considered; the technique is applicable to problems in which the axial thrust depends on the transverse deflection as well as problems with prescribed axial thrust. The formulas presented constitute the theory for a computer program of broad applicability, not only in the analysis of fuel rod bowing, but also to almost any multispan beam, particularly when the effects of axial loads cannot be neglected. 17 references. (NSA 22: 22866)

  10. Measurement of 2-dimensional local instantaneous liquid film thickness around simulated nuclear fuel rod by ultrasonic transmission technique

    International Nuclear Information System (INIS)

    An accurate knowledge of hydro-dynamic behavior of a liquid film flow on nuclear fuel rods is indispensable for analysis of the CHF under postulated loss-of-coolant-accidents in boiling water reactors. This work is concerned with a new development of ultrasonic transmission technique for film flow measurements. The technique adopted a rotating reflector, capable of measuring time-dependent spatial distribution of liquid film thickness around a simulated nuclear fuel rod. The scanning time is currently 4 ms for reconstruction of one image of the circumferential film thickness distribution. (orig.)

  11. Fuel performance improvement program: description and characterization of HBWR Series H-2, H-3, and H-4 test rods

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Barner, J.O.; Welty, R.K.

    1980-03-01

    The fabrication process and as-built characteristics of the HBWR Series H-2 and H-3 test rods, as well as the three packed-particle (sphere-pac) rods in HBWR Series H-4 are described. The HBWR Series H-2, H-3, and H-4 tests are part of the irradiation test program of the Fuel Performance Improvement Program. Fifteen rods were fabricated for the three test series. Rod designs include: (1) a reference dished pellet design incorporating chamfered edges, (2) a chamfered, annular pellet design combined with graphite-coated cladding, and (3) a sphere-pac design. Both the annular-coated and sphere-pac designs include internal pressurization using helium.

  12. Extended burnup demonstration: reactor fuel program. Pre-irradiation characterization and summary of pre-program poolside examinations. Big Rock Point extended burnup fuel

    International Nuclear Information System (INIS)

    This report is a resource document characterizing the 64 fuel rods being irradiated at the Big Rock Point reactor as part of the Extended Burnup Demonstration being sponsored jointly by the US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities. The program entails extending the exposure of standard BWR fuel to a discharge average of 38,000 MWD/MTU to demonstrate the feasibility of operating fuel of standard design to levels significantly above current limits. The fabrication characteristics of the Big Rock Point EBD fuel are presented along with measurement of rod length, rod diameter, pellet stack height, and fuel rod withdrawal force taken at poolside at burnups up to 26,200 MWD/MTU. A review of the fuel examination data indicates no performance characteristics which might restrict the continued irradiation of the fuel

  13. Statistical analysis in the design of nuclear fuel cells and training of a neural network to predict safety parameters for reactors BWR

    International Nuclear Information System (INIS)

    In this work the obtained results for a statistical analysis are shown, with the purpose of studying the performance of the fuel lattice, taking into account the frequency of the pins that were used. For this objective, different statistical distributions were used; one approximately to normal, another type X2 but in an inverse form and a random distribution. Also, the prediction of some parameters of the nuclear reactor in a fuel reload was made through a neuronal network, which was trained. The statistical analysis was made using the parameters of the fuel lattice, which was generated through three heuristic techniques: Ant Colony Optimization System, Neuronal Networks and a hybrid among Scatter Search and Path Re linking. The behavior of the local power peak factor was revised in the fuel lattice with the use of different frequencies of enrichment uranium pines, using the three techniques mentioned before, in the same way the infinite multiplication factor of neutrons was analyzed (k..), to determine within what range this factor in the reactor is. Taking into account all the information, which was obtained through the statistical analysis, a neuronal network was trained; that will help to predict the behavior of some parameters of the nuclear reactor, considering a fixed fuel reload with their respective control rods pattern. In the same way, the quality of the training was evaluated using different fuel lattices. The neuronal network learned to predict the next parameters: Shutdown Margin (SDM), the pin burn peaks for two different fuel batches, Thermal Limits and the Effective Neutron Multiplication Factor (keff). The results show that the fuel lattices in which the frequency, which the inverted form of the X2 distribution, was used revealed the best values of local power peak factor. Additionally it is shown that the performance of a fuel lattice could be enhanced controlling the frequency of the uranium enrichment rods and the variety of the gadolinium rods

  14. Simulation of the fuel rod bundle test QUENCH-03 using the integral code ASTEC V2

    Energy Technology Data Exchange (ETDEWEB)

    Kruse, Philipp; Koch, Marco K. [Bochum Univ. (Germany). Chair of Energy Systems and Energy Economics

    2010-05-15

    Failure of the main and emergency cooling-systems can lead to an accident with severe core degradation even with core meltdown. To prevent total meltdown of the uncovered and overheated core reflooding with water is an unavoidable accident management measure. The fast supply of water and the resulting increased available amount of steam can lead to crack formations and break up of the oxide layer of the fuel rods. The additionally exposed surface could result in an increased release of hydrogen due to a supplementary exothermal zirconium- steam-oxidation reaction. Within the frame of the QUENCH test-program - realised by FZK - loss of coolant accidents in LWR (Light Water Reactor) are analysed using an experimental reactor core to determine the produced amount of hydrogen, the so-called hydrogen source term. Additionally, the behaviour of the bundle with different absorber rod and cladding materials is being analysed. Based on the post-test calculations of the QUENCH tests with the severe accident code system ASTEC the capability of the code can be established and evaluated. In the following the post-test calculations of the QUENCH-03 test with ASTEC V2 are discussed. (orig.)

  15. Study of intermediate configurations during the fuel reload in BWRs; Estudio de configuraciones intermedias durante la recarga de combustible en BWR's

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia del C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Jacinto C, S., E-mail: luis.fuentes@inin.gob.mx [Universidad Autonoma del Estado de Yucatan, Calle 60 No. 491-A por 57, 97000 Merida, Yucatan (Mexico)

    2012-10-15

    The criticality state of the core of a boiling water reactor (BWR) was evaluated, during the reload process for the intermediate states between the load pattern of cycle end and the beginning of the next, using the information of the load pattern of the operation cycles 13 and 14 of Unit 1 of the nuclear power plant of Laguna Verde. For this evaluation the codes CASMO-4 and Simulate-3 for conditions of the core in cold were used. The strategy consisted on moving assemblies with 4 burned cycles of the reactor core. Later on were re situated the remaining assemblies, placing them in the positions to occupy in the next operation cycle. Finally, was carried out the assemblies load of fresh fuel. In each realized change, it was observing the behavior of the k-effective value that is the parameter used to evaluate the criticality state of each state of the core change. In a second stage, was designed a program that builds in automatic way each one of the intermediate cores and also analyzes the criticality state of the reactor core after each withdrawal, re situated and load of fuel assemblies. (Author)

  16. Two-dimensional thermal analysis of a fuel rod by finite volume method

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Rhayanne Y.N.; Silva, Mario A.B. da; Lira, Carlos A.B. de O., E-mail: ryncosta@gmail.com, E-mail: mabs500@gmail.com, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamaento de Energia Nuclear

    2015-07-01

    In a nuclear reactor, the amount of power generation is limited by thermal and physic limitations rather than by nuclear parameters. The operation of a reactor core, considering the best heat removal system, must take into account the fact that the temperatures of fuel and cladding shall not exceed safety limits anywhere in the core. If such considerations are not considered, damages in the fuel element may release huge quantities of radioactive materials in the coolant or even core meltdown. Thermal analyses for fuel rods are often accomplished by considering one-dimensional heat diffusion equation. The aim of this study is to develop the first paper to verify the temperature distribution for a two-dimensional heat transfer problem in an advanced reactor. The methodology is based on the Finite Volume Method (FVM), which considers a balance for the property of interest. The validation for such methodology is made by comparing numerical and analytical solutions. For the two-dimensional analysis, the results indicate that the temperature profile agree with expected physical considerations, providing quantitative information for the development of advanced reactors. (author)

  17. Brazing of AM-350 stainless steel LWBR fuel rod support grids (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Ebejer, L.P.

    1979-02-01

    A brazing process has been developed wherein several hundred stamped AM-350 stainless steel sheet metal components, wire components and machined bar components were simultaneously joined together to fabricate about 400 grids of different sizes for the LWBR fuel rod support system. High temperature (2110F +- 20F) vacuum brazing was employed using Ni--Cr--Si braze alloy filler metal in the form of paste. Techniques employed in the assembly, braze alloy application and fixturing of grids to achieve adequate dimensional control are discussed in detail. The brazing thermal cycle as related to the complex metallurgical process of both AM-350 stainless steel and the Ni--Cr--Si braze alloy is also discussed.

  18. CFD analysis of coolant channel geometries for a tightly packed fuel rods assembly of Super FBR

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Rui, E-mail: guorui@asagi.waseda.jp; Oka, Yoshiaki

    2015-07-15

    Highlights: • Supercritical water heat transfer is validated against the experimental data. • Thermal hydraulic performance of three coolant channel cross-sectional geometries are analyzed. • Geometry B is superior to the other two geometries. - Abstract: This paper presents the CFD investigation on three cross-sectional geometries of coolant channels in the newly designed tightly packed fuel rods assembly for high breeding of a supercritical-pressure light water cooled fast breeding reactor (Super FBR). The calculations addressed the turbulence models and mesh conditions validated against the experimental data. The cladding temperatures and pressure drop are compared. It is concluded that geometry B (triangle with round corner) is superior to the other two (round and triangle with sharp corner) due to its excellent thermal hydraulic characteristics.

  19. Sliding Wear and Friction Behavior of Fuel Rod Material in Water and Dry State

    Science.gov (United States)

    Park, Jin Moo; Kim, Jae Hoon; Jeon, Kyeong Lak; Park, Jun Kyu

    In water cooled reactors, the friction between spacer grid and fuel rod can lead to severe wear and it is an important topic to study. In the present study, sliding wear behavior of zirconium alloy was investigated in water and dry state using the pin-on-disc sliding wear tester. Sliding wear resistance of zirconium alloy against heat treated inconel alloy was examined at room temperature. The parameters in this study were sliding velocity, axial load and sliding distance. The wear characteristics of zirconium alloy was evaluated by friction coefficient, specific wear rate and wear volume. The micro-mechanisms responsible for wear in zirconium alloy were identified to be micro-cutting, micro-pitting, delamination and micro-cracking of deformed surface zone.

  20. Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

    International Nuclear Information System (INIS)

    During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

  1. Analysis of neutron parameters in light water moderated lattices of ThO2 and UO2 fuel rods

    International Nuclear Information System (INIS)

    A large number of light water moderated lattices of UO2 and ThO2 fuel rods were analyzed with the code HAMMER. The purpose of the study was to compare experimental results with computer calculated values. The model employed is described and some modification were introduced in the resonance parameters of Th-232 to increase the agreement with the experimental value

  2. Tomography on nuclear fuel rods in the nuclear power plant of Dodewaard. Tomografie aan splijtstofstaven in de centrale Dodewaard

    Energy Technology Data Exchange (ETDEWEB)

    Tanke, R.H.J.; Jaspers, J.E.; Gaalman, P.A.M. (KEMA, Arnhem (Netherlands). Division Research and Development)

    1990-09-06

    This report discusses the feasibility of using emission tomography on fuel rods in the Dodewaard reactor. The tomography can be used to increase the efficiency of the use of fissionable material. (R.A.B.). 4 refs.; 17 figs.; 1 tab.

  3. Starting point, keys and milestones of a computer code for the simulation of the behavior a nuclear fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A.C. [CNEA, Centro Atomico Bariloche, CAB, 8400 S. C. de Bariloche (Argentina)]. e-mail: marino@cab.cnea.gov.ar

    2007-07-01

    The BaCo code ('Barra Combustible') was developed at the Atomic Energy National Commission of Argentina (CNEA) for the simulation of the behavior of a nuclear fuel rod under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement and validation of a BaCo during the last 30 years. 'Extreme case analysis', parametric (or sensibility), probabilistic (or statistic) analysis plus the analysis of the fuel performance (full core analysis) are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP and the design of advanced fuel elements as CARA and CAREM .The 3D additional tools of BaCo can enhance the understanding of the fuel rod behavior, the fuel design and the safety margins. The modular structure of the BACO code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity. (Author)

  4. A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

    Directory of Open Access Journals (Sweden)

    Martina Adorni

    2011-01-01

    Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

  5. Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

    Directory of Open Access Journals (Sweden)

    Armando C. Marino

    2011-01-01

    Full Text Available The BaCo code (“Barra Combustible” was developed at the Atomic Energy National Commission of Argentina (CNEA for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity, probabilistic (or statistic analysis plus the analysis of the fuel performance (full core analysis are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

  6. Post-irradiation examination of prototype Al-64 wt% U3Si2 fuel rods from NRU

    International Nuclear Information System (INIS)

    Three prototype fuel rods containing Al-64 wt% U3Si2 (3.15 gU/cm3) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U3Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U3Si2 powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37 degrees C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U3Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL's research reactors

  7. Operational model for the safety of defective spent fuel rods in dry storage

    International Nuclear Information System (INIS)

    In dry storage, the breach of the container and cladding will expose the spent fuel to oxygen. This incidental scenario could be detrimental for the integrity of the spent fuel rod. Indeed, UO2 is not thermodynamically stable in air. The oxidation of UO2 up to U3O8 leads to crystalline swelling and sample bulking that may increase the initial defect in the cladding and enhance the release of fission products. Oxidation experiments with in-situ characterisation of structural phases have been performed on unirradiated UO2 powders and led to propose an oxidation model for the first step of oxidation, i.e. from UO2 to U3O7 (Poulesquen et al., 2007). This model has been up-graded in order to describe the oxidation of spent fuel fragments. It takes into account the stoichiometric phases, which have observed during spent fuel oxidation, and the role of grain boundaries. The oxidation model is able to reproduce the different weight gain curves provided in literature and obtained as part of the French PRECCI program. Moreover, the diffusion coefficients of oxygen in the intermediate structural phases, which have been derived from the fitting of the model to experimental data, are consistent with literature data. Furthermore in-situ SEM investigations during oxidation of UO2 grain have evidenced sample bulking, due to grain cracking, as soon as a critical thickness of U3O7 is achieved in the grain. It has led to propose a new criterion for spent fuel degradation. The time to achieve this criterion is calculated by using the oxidation model. A duration of safety, or time before consequent fracturing of spent fuel in air, is thus deduced. Sensitivity and uncertainty analyses allow us to express the duration of safety as a simplified analytical formula of the input parameters of the oxidation model, such as temperature of storage and spent fuel characteristics. The duration of safety can be qualitatively linked to irradiation parameters such as the burn up, the linear power or

  8. A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel

    OpenAIRE

    Jacobsson, Staffan; Håkansson, Ane; Andersson, Camilla; Jansson, Peter; Bäcklin, Anders

    1998-01-01

    A tomographic method for experimental investigation of the integrity of usedLWR fuel has been developed. It is based on measurements of the gamma radiation fromthe fission products in the fuel rods. A reconstruction code of the algebraic type has beenwritten. The potential of the technique has been examined in extensive simulationsassuming a gamma-ray energy of either 0.66 MeV (137Cs) or 1.27 MeV (154Eu).The resultsof the simulations for BWR fuel indicate that single fuel rods or groups of ro...

  9. Effect of applied current on the formation of defect in PWR nuclear fuel rods in resistance pressure welding process

    International Nuclear Information System (INIS)

    The welding of zirconium alloy components is one of the most critical processes in the fabrication of nuclear fuel rods used in pressurized water reactors. For this, various welding processes, such as gas tungsten arc welding, electron beam welding, laser beam welding, and resistance pressure welding (RPW), are used around the world. In Korea, the RPW process is being used to fabricate nuclear fuel assembly fuel rods. This study investigated changes in the weldment shape owing to welding conditions such as welding current, welding force, and overlapping. The welding soundness of the weldment was evaluated by hydraulic burst test. The welding temperature of the weld zone was measured using a thermal infrared method. Discontinuous black spots in the weld line, regarded as a non-bonding defect, were confirmed as spots caused by the carbide precipitation of zirconium during welding. (author)

  10. Neutron Flux Depression in the UO2-PuO2 (15 to 30%) Fuel Rods from IVO-FR2-Vg7-Irradiation Experiment

    International Nuclear Information System (INIS)

    The thermal-neutron flux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO2-PUO2 (15 to 30% PUO2) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (Author) 22 refs

  11. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  12. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    Energy Technology Data Exchange (ETDEWEB)

    Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL, will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT.

  13. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    Energy Technology Data Exchange (ETDEWEB)

    Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

  14. Feasibility evaluation of x-ray imaging for measurement of fuel rod bowing in CFTL test bundles

    International Nuclear Information System (INIS)

    The Core Flow Test Loop (CFTL) is a high temperature, high pressure, out-of-reactor helium-circulating system. It is designed for detailed study of the thermomechanical performance, at prototypic steady-state and transient operating conditions, of electrically heated rods that simulate segments of core assemblies in the Gas-Cooled Fast Breeder reactor demonstration plant. Results are presented of a feasibility evaluation of x-ray imaging for making measurements of the displacement (bowing) of fuel rods in CFTL test bundles containing electrically heated rods. A mock-up of a representative CFTL test section consisting of a test bundle and associated piping was fabricated to assist in this evaluation

  15. Characterization of a suspect nuclear fuel rod in a case of illegal international traffic of fissile material.

    Science.gov (United States)

    Capannesi, G; Vicini, C; Rosada, A; Avino, P

    2010-06-15

    This case study describes the characterization of a suspect rod of nuclear fuel seized in Italy: on request of the coroner, the characterization concerned the kind and the conditions of the rod, the amount and the specific characteristics of the species present in it, with particular attention to their possible use chemical and/or nuclear plants. The methodology used was based on radiochemical analyses (gammagraphic and gamma-spectrometry) whereas the comparison was performed by means of a fuel reference element working in the TRIGA nuclear reactor at Research Center of ENEA-Casaccia. The results show clearly how the exhibit was an element of nuclear fuel, how long it was irradiated, and the amount of (239)Pu produced and the (235)U consumed. Finally, even if the seized rod was briefly radiated at the "zero power" and traces of fission products and plutonium were found, it would be still usable as "fresh" fuel in a reactor type TRIGA if it had not been intercepted by Italian police authorities.

  16. An investigation towards real time dose rate monitoring, and fuel rod detection in a First Generation Magnox Storage Pond (FGMSP).

    Science.gov (United States)

    Jackson, Sarah F; Monk, Stephen D; Riaz, Zahid

    2014-12-01

    The First Generation Magnox Storage Pond (FGMSP) is located on the Sellafield Nuclear Site, housing legacy spent Magnox nuclear fuel. Some of which has since corroded, forming a layer of Corroded Magnox Sludge (CMS) creating one of the largest decommissioning challenges the UK has faced. In this work the composition, physical properties and potentially high hazard nature of CMS are discussed, as are the gamma emission spectra of spent Magnox fuel rods typical of the ilk stored. We assess the potential use of a RadLine gamma detector to dose rate map this area and provide fuel rod detection. RadLine consists of a small scintillator, fibre optic cable and photon counter. The probe has the unusual advantage of not being electrically active and therefore fully submersible underwater, with the option to deploy hundreds of metres in length. Our experimental method encompasses general purpose Monte Carlo radiation transport code, MCNP, where we describe the modelling of CMS and pond liquor in comprehensive detail, including their radiological spectrum, chemical composition data, and physical properties. This investigation concludes that the maximum energy deposited within the scintillator crystal due to ambient CMS corresponds to a dose rate of 5.65Gy h(-1), thus above this value positive detection of a fuel rod would be anticipated. It is additionally established that the detectable region is within a 20cm range. PMID:25244071

  17. Control rod effects on reaction rate distributions in tight pitched PuO2-UO2 fuel assembly

    International Nuclear Information System (INIS)

    Investigations were made for the heterogeneity effects caused by insertion or withdrawal of a B4C control rod on fine structure of reaction rates distributions in a tight pitched PuO2-UO2 fuel assembly. Analysis was carried out by using the VIM and SRAC codes with the libraries based on JENDL-2 for the hexagonal fuel assembly basically corresponding to the PROTEUS-LWHCR experimental core. The reaction rates are affected more remarkably by the withdrawal of the control rod rather than its insertion. The changes of the reaction rates were decomposed into three terms of spectrum shifts, the changes of effective cross sections with fine groups, and their higher order components. From the analysis, it is concluded that most changes of reaction rates are caused by spectral shifts. The SRAC code with fine group constants can predict the distribution of reaction rates and their ratios with the accuracy of about 5 % except for the values related to Pu-242 capture rate, as compared with the VIM results. To increase the accuracy, it is necessary to generate the effective cross sections of the fuel near control rods with consideration of the heterogeneities in the fuel assembly. (author)

  18. Characterization of a suspect nuclear fuel rod in a case of illegal international traffic of fissile material.

    Science.gov (United States)

    Capannesi, G; Vicini, C; Rosada, A; Avino, P

    2010-06-15

    This case study describes the characterization of a suspect rod of nuclear fuel seized in Italy: on request of the coroner, the characterization concerned the kind and the conditions of the rod, the amount and the specific characteristics of the species present in it, with particular attention to their possible use chemical and/or nuclear plants. The methodology used was based on radiochemical analyses (gammagraphic and gamma-spectrometry) whereas the comparison was performed by means of a fuel reference element working in the TRIGA nuclear reactor at Research Center of ENEA-Casaccia. The results show clearly how the exhibit was an element of nuclear fuel, how long it was irradiated, and the amount of (239)Pu produced and the (235)U consumed. Finally, even if the seized rod was briefly radiated at the "zero power" and traces of fission products and plutonium were found, it would be still usable as "fresh" fuel in a reactor type TRIGA if it had not been intercepted by Italian police authorities. PMID:20223608

  19. ROBOT3: a computer program to calculate the in-pile three-dimensional bowing of cylindrical fuel rods (AWBA Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Kovscek, S.E.; Martin, S.E.

    1982-10-01

    ROBOT3 is a FORTRAN computer program which is used in conjunction with the CYGRO5 computer program to calculate the time-dependent inelastic bowing of a fuel rod using an incremental finite element method. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the CYGRO5 axisymmetric model. Fuel rod supports are modeled as displacement, force, or spring-type nodal boundary conditions. The program input is described and a sample problem is given.

  20. Application of STAV5 code for the analysis of fission gas release in power reactor rods

    International Nuclear Information System (INIS)

    STAV5 is a design code for calculation of temperatures, fission gas release and rod pressure in BWR and PWR fuel rods. It includes the treatment of pellet cracks affecting conductivity and thermal expansion, gap closure by eccentric or relocated pellet fragments and oxide and crud build-up on the clad outer surface. The fission gas release model consists of two parts: High temperature release based on grain boundary saturation and low temperature release varying with fission rate of different isotopes. STAV5 has been benchmarked with a number of inpile thermal measurement experiments to as high burnup as 25 MWd/kg U. The main application of STAV5 is as a routine design tool for power reactor rods. It is also used to compare with PIE data. Examples are given from the analyses of fission gas release data from BWR rods from Oskarshamn 1 and Barsebeck 1 as well as PWR rods from Maine Yankee initial cores. The STAV5 evaluation show the importance of power histories, densification and the position in the assembly. (author)

  1. Physical models and codes for prediction of activity release from defective fuel rods under operation conditions and in leakage tests during refuelling

    International Nuclear Information System (INIS)

    It is appropriate to use the dependences, based on physical models, in the design-analytical codes for improving of reliability of defective fuel rod detection and for determination of defect characteristics by activity measuring in the primary coolant. In the paper the results on development of some physical models and integral mechanistic codes, assigned for prediction of defective fuel rod behaviour are presented. The analysis of mass transfer and mass exchange between fuel rod and coolant showed that the rates of these processes depends on many factors, such as coolant turbulent flow, pressure, effective hydraulic diameter of defect, fuel rod geometric parameters. The models, which describe these dependences, have been created. The models of thermomechanical fuel behaviour, stable gaseous FP release were modified and new computer code RTOP-CA was created thereupon for description of defective fuel rod behaviour and activity release into the primary coolant. The model of fuel oxidation in in-pile conditions, which includes radiolysis and RTOP-LT after validation of physical models are planned to be used for prediction of defective fuel rods behaviour

  2. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    Science.gov (United States)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  3. Power Burst Facility: U(18)O2-CaO-ZrO2 Fuel Rods in Water

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Alexis Weir; Valeria L. Putnam; John D. Bess

    2009-09-01

    The Power Burst Facility (PBF) reactor operated from 1972 to 1985 on the SPERT Area I of the Idaho National Laboratory, then known as Nuclear Reactor Test Station. PBF was designed to provide experimental data to aid in defining thresholds for and modes of failure under postulated accident conditions. PBF reactor startup testing began in 1972. This evaluation focuses on two operational loading tests, chronologically numbered 1 and 2, published in a startup-test report in 1974 [1]. Data for these tests was used by one of the authors to validate a MCNP model for criticality safety purposes [2]. Although specific references to original documents are kept in the text, all the reactor parameters and test specific data presented here was adapted from that report. The tests were performed with operational fuel loadings, a stainless steel in-pile tube (IPT) mockup, a neutron source, four pulse chambers, two fission chambers, and one ion chamber. The reactor's four transition rods (TRs) and control rods (CRs) were present but TR boron was completely withdrawn below the core and CR boron was partially withdrawn above the core. Test configurations differ primarily in the number of shim rods, and consequently the number of fuel rods included in the core. The critical condition was approached by incrementally and uniformly withdrawing CR boron from the core. Based on the analysis of the experimental data and numerical calculations, both experiments are considered acceptable as criticality safety benchmarks.

  4. Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

    Science.gov (United States)

    Wang, Hong; Wang, Jy-An John

    2016-10-01

    Behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending was studied. Tests were performed under load or moment control at 5 Hz. The surrogate rods fractured under moment amplitudes greater than 10.16 Nm with fatigue lives between 2.4 × 103 and 2.2 × 106 cycles. Fatigue response of Zry-4 cladding was characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition affect surrogate rod failure. Both debonding of PPI/PCI and pellet fracturing contribute to surrogate rod bending fatigue. The effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective gauge length is effective in sensor spacing correction. The database developed and the understanding gained in this study can serve as input to analysis of SNF (spent nuclear fuel) vibration integrity.

  5. Leaking Fuel Impacts and Practices

    International Nuclear Information System (INIS)

    The impact of leaking fuel rods on the operation of nuclear power plants and the practices of handling leaking fuel has been reviewed by the CSNI Working Group on Fuel Safety in order to promote a better understanding on the handling of leaking fuel in power reactors, as well as to discuss and review the current practices in member countries to help in decisions on the specification of reactor operation conditions with leaking fuel rods and on the handling of leaking fuel after removal from reactor. Experts from 15 countries provided data on the handling of leaking fuel in PWR, BWR, VVER and PHWR reactor types. The review covered the operation of NPP reactors with leaking fuel, wet and dry storage and transport of leaking assemblies. The methods and applied instruments to identify leaking fuel assemblies and the repair of them were addressed in the review. Special attention was paid to the activity release from leaking rods in the reactor and under storage conditions. The consideration of leaking fuel in safety analyses on core behaviour during postulated accidents was also discussed in the review. The main conclusions of the review pointed out that the activity release from leaking fuel rods in the reactor can be handled by technological systems, or in case of failure of too many rods the reactor can be shutdown to minimize activity release. Under accident conditions and operational transients the leaking rods may produce coolant activity concentration peaks. The storage of spent leaking fuel is normally characterised by moderate release of radionuclides from the fuel. The power plants apply limits for activity concentration to limit the amount of leaking rods in the core. In different countries, the accident analyses take into consideration the potential release from leaking fuel rods in design basis accidents in different ways. Some power plants apply special tools for handling and repair of leaking assemblies and rods. The leaking rods are stored together with

  6. Nuclear coupled flow instability study for natural circulation BWR startup transient

    International Nuclear Information System (INIS)

    Natural circulation Boiling Water Reactor (BWR) startup transient was investigated in Purdue University Multidimensional Test Assembly (PUMA) facility based on a natural circulation BWR design. Strategy and results of the experiments, which consider the effects of void-reactivity and fuel heat conduction time constant, are discussed. Total reactivity is treated to be composed of two components: external reactivity due to control rod motion and void-reactivity. A detailed analysis for heat conduction problem is performed to derive dimensionless groups. Based on area-averaged heat conduction equations for pellet and clad regions, Fourier and Biot numbers are derived to simulate wall heat flux response. Power transient, which has been used for startup transient investigation without void-reactivity feedback is used to derive the control rod reactivity. Twelve conductivity probes are used to measure local void fraction inside core at three axial locations. The local void-fraction data is used to calculate volume average void fraction, which is used to calculate the voil-reactivity. A real-time Point Kinetic Model solver is implemented to PUMA heater power control program to determine power transient during startup. The results demonstrate that the inclusion of void-reactivity feedback worsen the scenario for startup instabilities and may cause large amplitude neutron flux oscillations. (author)

  7. Specialists' meeting on LMFBR fuel rod behaviour under operational transients, Kalpakkam, India, 3-6 December 1985

    International Nuclear Information System (INIS)

    IWGFR Specialists' Meeting on ''LMFBR Fuel Rod Behaviour Under Operational Transients'' was held in Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 3-6 December 1985. The meeting was attended by the representatives of Belgium, France, FRG, India, Italy, US, USSR, IAEA and observers from India. The purpose of the meeting was to provide a forum for the exchange of information on the subject of LMFBR Fuel Rod Behaviour Under Operational Transients. The meeting presentations were divided into sessions devoted to the following topics: Overview of National Programmes (3 papers); In-pile and out-pile experimental facilities and results (5 papers); Modelling and Code work (5 papers). A separate abstract was prepared for each of these papers

  8. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro, E-mail: soba@cnea.gov.ar; Denis, Alicia

    2015-10-15

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  9. Study of film thickness on fuel rod under high pressure and high temperature steam-water two phase flow

    International Nuclear Information System (INIS)

    The liquid film thickness on the fuel rod is measured by ultrasonic echo technique under high temperature and high pressure steam-water two phase flow. As quality is increased, film thickness is decreased. The film thickness is about 0.2 mm at 9% of quality under 1 MPa. It was found from test data that disturbance wave is not measured clearly and change of film is small. (author)

  10. Flow-Induced Vibration Measurement of an Inner Cladding Tube in a Simulated Dual-Cooled Fuel Rod

    International Nuclear Information System (INIS)

    To create an internal coolant flow passage in a dual cooled fuel rod, an inner cladding tube cannot have intermediate supports enough to relieve its vibration. Thus it can be suffered from a flow-induced vibration (FIV) more severely than an outer cladding tube which will be supported by series of spacer grids. It may cause a fatigue failure at welding joints on the cladding's end plug or fluid elastic instability of long, slender inner cladding due to decrease of a critical flow velocity. This is one of the challenging technical issues when a dual cooled fuel assembly is to be realized into a conventional reactor core To study an actual vibration phenomenon of a dual cooled fuel rod, FIV tests using a small-scale test bundle are being carried out. Measurement results of inner cladding tube of two typically simulated rods are presented. Causes of the differences in the vibration amplitude and response spectrum of the inner cladding tube in terms of intermediate support condition and pellet stacking are discussed

  11. Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor

    International Nuclear Information System (INIS)

    The composition calculation and the activity of nuclear materials subject to processes of burnt, irradiation and decay periods are of utility for diverse activities inside the nuclear industry, as they are it: the processes design and operations that manage radioactive material, the calculation of the inventory and activity of a core of burnt nuclear fuel, for studies of type Probabilistic Safety Analysis (APS), as well as for regulation processes and licensing of nuclear facilities. ORIGEN is a program for computer that calculates the composition and the activity of nuclear materials subject to periods of burnt, irradiation and decay. ORIGEN generates a great quantity of information whose processing and analysis are laborious, and it requires thoroughness to avoid errors. The automation of the extraction, conditioning and classification of that information is of great utility for the analyst. By means of the use of the post-processor presented in this work it is facilitated, it speeds up and wide the capacity of analysis of results, since diverse consultations with several classification options and filtrate of results can be made. As illustration of the utility of the post-processor, and as an analysis of interest for itself, it is also presented in this work the composition of the activity of a burned core in a BWR type reactor according to the following classification criteria: by type of radioisotope (fission products, activation products and actinides), by specie type (gassy, volatile, semi-volatile and not volatile), by element and by chemical group. The results show that the total activity of the studied core is dominated by the fission products and for the actinides, in proportion four to one, and that the gassy and volatile species conform a fifth part of the total activity of the core. (Author)

  12. Advances in Forecasting and Prevention of Resonances Between Coolant Acoustical Oscillations and Fuel Rod Vibrations

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich [NPP, NPEI, 14, Krasnokazarmennaya str. Moscow, 111250 (Russian Federation)

    2009-06-15

    To prevent the appearance of the conditions for resonance interaction between the fluid flow and the reactor internals (RI), fuel rod (FR ) and fuel assemblies (FA) it is necessary to de-tune Eigen frequency of coolant pressure oscillations (EFCPO) and natural frequency of mechanical element's oscillations and also of the system which is formed by the comprising of these elements. Other words it is necessary to de-tune acoustic resonance frequency and natural frequencies of RI, FR and FA. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of the coolant outside of which there is no resonant interaction with structure vibrations. The presented work is devoted to finding the solution of this problem. There are results of an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. Abnormal growth of intensity of pressure pulsations in a mode with definite value of reactor capacity have been found out by measurements on VVER - 1000 reactor. This phenomenon has been found out casually and its original reason had not been identified. Paper shows that disappearance of this effect could be reached by realizing outlet of EFCPO from so-called, pass bands of frequencies (PBF). PBF is located symmetrical on both parties from frequency of own oscillations of FA. Methods, algorithms of calculations and quantitative estimations are developed for EFCPO, Q and PBF in various modes of operation NPP with VVER-1000. Results of calculations allow specifying area of resonant interaction EFCPO with vibrations of FR, FA and a basket of reactor core. For practical realization of the received results it is offered to make corresponding additions to the design documentation and maintenance instructions of the equipment of the NPP with VVER-1000. The improvement of these documents

  13. Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

    Science.gov (United States)

    Solonin, V. I.; Perevezentsev, V. V.

    2012-05-01

    Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

  14. Consolidation of spent fuel rods from LWRs: Current procedures and future plans with reference to work in the USA and Germany

    International Nuclear Information System (INIS)

    This report presents the current status of the process by which light water reactor spent fuel is consolidated for the purpose of decreasing the space needed for interim storage and final disposal. The process is called rod consolidation. Light water reactor fuel consists of two distinct sets of components: the fuel rods and the skeleton (spacer grids, end-fittings and guide tubes), termed non-fuel bearing components (NFBCs). The process entails the disassembly of irradiated nuclear fuel into individual fuel rods and the NFBCs. Fuel compaction ratios of 2 to 1 can be achieved, thus the rods from two fuel assemblies can be stored in the space required for one assembly. NFBCs compaction ratios of 8 or 10 to 1 can be achieved, thus the skeletons from 8 or 10 fuel assemblies can be stored in the space required for a single assembly. The information presented in this report is intended to provide a basis for incorporation of the rod consolidation process into an effective overall spent fuel management strategy (interim storage to final disposal). Since rod consolidation is a process, it must be performed in a facility. This can be either a facility dedicated to the process (e.g. away-from-reactor hot cell) or an existing facility (e.g. at-reactor spent fuel pool). The report is also intended to acquaint the reader with the many rod consolidation concepts. The report aims at developing an understanding of the fundamentals of rod consolidation and appreciation of the interdependency of the process and the facility in which it will take place. It can also be of interest to those who decide on constructing additional capacity or even transshipment. The reason is that rod consolidation will result in lowering the number of handling events because each handling event will move twice as much fuel. Lower handling is desirable because it is the main cause of abnormal incidents in spent fuel management. Also for shipment, the number of casks is reduced, lowering the risk to the

  15. Spent fuel disassembly hardware and other non-fuel bearing components: characterization, disposal cost estimates, and proposed repository acceptance requirements

    International Nuclear Information System (INIS)

    There are two categories of waste considered in this report. The first is the spent fuel disassembly (SFD) hardware. This consists of the hardware remaining after the fuel pins have been removed from the fuel assembly. This includes end fittings, spacer grids, water rods (BWR) or guide tubes (PWR) as appropriate, and assorted springs, fasteners, etc. The second category is other non-fuel-bearing (NFB) components the DOE has agreed to accept for disposal, such as control rods, fuel channels, etc., under Appendix E of the standard utiltiy contract (10 CFR 961). It is estimated that there will be approximately 150 kg of SFD and NFB waste per average metric ton of uranium (MTU) of spent uranium. PWR fuel accounts for approximately two-thirds of the average spent-fuel mass but only 50 kg of the SFD and NFB waste, with most of that being spent fuel disassembly hardware. BWR fuel accounts for one-third of the average spent-fuel mass and the remaining 100 kg of the waste. The relatively large contribution of waste hardware in BWR fuel, will be non-fuel-bearing components, primarily consisting of the fuel channels. Chapters are devoted to a description of spent fuel disassembly hardware and non-fuel assembly components, characterization of activated components, disposal considerations (regulatory requirements, economic analysis, and projected annual waste quantities), and proposed acceptance requirements for spent fuel disassembly hardware and other non-fuel assembly components at a geologic repository. The economic analysis indicates that there is a large incentive for volume reduction

  16. Use of high energy gamma emission tomography for partial defect verification of spent fuel assemblies

    International Nuclear Information System (INIS)

    The possibility to use passive gamma emission tomography for revealing non-destructively the rod structure of spent BWR fuel assemblies has been studied in cooperation with the Finnish Support Programme to the IAEA Safeguards (task FIN A98) and the Technical University of Budabest in Hungary. The ultimate goal is to develop partial verification methods for verification of spent nuclear fuel. The task included experimental measurements of irradiated BWR assemblies using underwater measurement techniques together with computer analysis of the measured data as well as computer simulation of tomographic measurements. The results obtained show that rod-level partial defect verification of spent LWR fuel assemblies is feasible using computed gamma emission tomography. This report describes the results of this project. (orig.). (7 refs., 29 figs., 2 tabs.)

  17. OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-02-01

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  18. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  19. DIONISIO 2.0: new version of the code for simulating the behavior of a power fuel rod under irradiation

    International Nuclear Information System (INIS)

    During the latest ten years the Codes and Models Section of the Nuclear Fuel Cycle Department has been developing the DIONISIO code, which simulates most of the main phenomena that take place within a fuel rod during the normal operation of a nuclear reactor: temperature distribution, thermal expansion, elastic and plastic strain, creep, irradiation growth, pellet-cladding mechanical interaction, fission gas release, swelling and densification. Axial symmetry is assumed and cylindrical finite elements are used to discretized the domain. The code has a modular structure and contains more than 40 interconnected models. A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO2 fuels in LWR conditions, predict the radial distribution of power density, burnup and concentration of diverse nuclides within the pellet. New models of porosity and fission gas release in the rim, as well as the influence of the microstructure of this zone on the thermal conductivity of the pellet, are presently under development. A considerable computational challenge was the inclusion of the option of simulating the whole bar, by dividing it in a number of axial segments, at the user's choice, and solving in each segment the complete problem. All the general rod parameters (pressure, fission gas release, volume, etc.) are evaluated at the end of every time step. This modification allows taking into account the axial variation of the linear power and, consequently, evaluating the dependence of all the significant rod parameters with that coordinate. DIONISIO was elected for participating in the FUMEX III project of codes intercomparison, organized by IAEA, from 2008 to 2011. The results of the simulations performed within this project were compared with more than 30 experiments that involve more than 150 irradiated rods. The high number

  20. LWR MOX fuel irradiation tests. HBWR irradiation with the instrument rig, IFA-514/565

    International Nuclear Information System (INIS)

    IFA-514 irradiation test was performed in Halden Reactor (HBWR) in Norway to study the irradiation performance of LWR MOX fuels. The fuel specifications for this irradiation test were decided in accordance with those of BWR 8x8 fuels, and plutonium content of MOX fuels was set to be 5.8 wt.%. Six MOX fuel rods, of which parameters were pellet geometry (solid or annular) and surface roughness (grinded or as-sintered), were irradiated to the assembly average burn-up of ∼45 GWd/t, and the instrument data during irradiation, i.e. cladding elongation, fuel stack elongation, fuel center temperature, and rod inner pressure, were obtained, and subsequent post-irradiation examinations provided the following results: No remarkable corrosion and deformation was observed on the irradiated fuel rods. The averaged irradiation growth was 0.13%, no nodular corrosion was observed, and the maximum thickness of oxide layer was ∼60 μm. The FP gas release rate was ∼21%. The irradiation for three of six fuel rods irradiated in IFA-514 irradiation tests were continued to the assembly average burn-up of ∼56 GWd/t in IFA-565 irradiation tests, and the results of this irradiation test were as follows: No remarkable corrosion and deformation was observed on the irradiated fuel rods. The FP gas release behavior of LWR MOX fuels was similar to that of BWR UO2 and ATR MOX fuels, and no difference was confirmed in the FP gas release behavior. Also FP gas release rate of annular pellets (∼13%) was lower than that of solid ones (∼16%). No remarkable difference in the effect of pellet geometry, i.e. solid and annular, on PCMI behavior was observed in any fuel rod from the instrument data of fuel stack elongation and the results of ceramography. However, the reduction of cladding diameter change in pellet geometry is expected since the cladding diameter change of annular fuel rods was less than that of solid ones. The disappearance of as-fabricated granular boundaries during irradiation

  1. On-site gamma-ray spectroscopic measurements of fission gas release in irradiated nuclear fuel.

    Science.gov (United States)

    Matsson, I; Grapengiesser, B; Andersson, B

    2007-01-01

    An experimental, non-destructive in-pool, method for measuring fission gas release (FGR) in irradiated nuclear fuel has been developed. Using the method, a significant number of experiments have been performed in-pool at several nuclear power plants of the BWR type. The method utilises the 514 keV gamma-radiation from the gaseous fission product (85)Kr captured in the fuel rod plenum volume. A submergible measuring device (LOKET) consisting of an HPGe-detector and a collimator system was utilised allowing for single rod measurements on virtually all types of BWR fuel. A FGR database covering a wide range of burn-ups (up to average rod burn-up well above 60 MWd/kgU), irradiation history, fuel rod position in cross section and fuel designs has been compiled and used for computer code benchmarking, fuel performance analysis and feedback to reactor operators. Measurements clearly indicate the low FGR in more modern fuel designs in comparison to older fuel types.

  2. 44-BWR WASTE PACKAGE LOADING CURVE EVALUATION

    International Nuclear Information System (INIS)

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an application of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing BWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results of 100 percent of the current BWR projected waste stream being able to be disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates is contingent upon the referenced waste stream being sufficiently similar to the waste stream received for disposal. (3) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials

  3. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

  4. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    Science.gov (United States)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  5. Modeling of the thermo-mechanical behaviour of the PWR fuel

    International Nuclear Information System (INIS)

    This article reviews the various physical phenomena that take place in an irradiated fuel rod and presents the development of the thermo-mechanical codes able to simulate them. Though technically simple the fuel rod is the place where appear 4 types of process: thermal, gas behaviour, mechanical and corrosion that combine involving 5 elements: the fuel pellet, the fuel clad, the fuel-clad gap, the inside volume and the coolant. For instance the pellet is the place where the following mechanical processes took place: thermal dilatation, elastic deformation, creep deformation, densification, solid swelling, gaseous swelling and cracking. The first industrial code simulating the behaviour of the fuel rod was COCCINEL, it was developed by AREVA teams from the American PAD code that was included in the Westinghouse license. Today the GALILEO code has replaced the COPERNIC code that was developed in the beginning of the 2000 years. GALILEO is a synthesis of the state of the art of the different models used in the codes validated for PWR and BWR. GALILEO has been validated on more than 1500 fuel rods concerning PWR, BWR and specific reactors like Siloe, Osiris, HFR, Halden, Studsvik, BR2/3,...) and also for extended burn-ups. (A.C.)

  6. Experience of complex use of radioactive methods of FFD on operating and shut down reactor for diagnostics of types of untight fuel rods

    International Nuclear Information System (INIS)

    The report contains a review of principles and methods, used by the experts of the Russian Research Center 'Kurchatov Institute' at diagnostics of fuel rods tightness during operation period of WWER type NPP, and also interpretation practices of obtained assessments. The agreement between assessments of untight rods parameters, which were determined by FFD radioactive methods applied for operating/shut down reactor, and the scale and character of rod leakage having a place in a reality is illustrated with concrete examples, taken from experience of WWER operation. (authors)

  7. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    International Nuclear Information System (INIS)

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s−1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  8. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Science.gov (United States)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  9. Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

    International Nuclear Information System (INIS)

    The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

  10. A comparison between genetic algorithms and neural networks for optimizing fuel recharges in BWR; Una comparacion entre algoritmos geneticos y redes neuronales para optimizar recargas de combustible en BWR's

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz J, J. [Instituto Nacional de Investigaciones Nucleares, Depto. Sistemas Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico); Requena, I. [Universidad de Granada (Spain)

    2002-07-01

    In this work the results of a genetic algorithm (AG) and a neural recurrent multi state network (RNRME) for optimizing the fuel reload of 5 cycles of the Laguna Verde nuclear power plant (CNLV) are presented. The fuel reload obtained by both methods are compared and it was observed that the RNRME creates better fuel distributions that the AG. Moreover a comparison of the utility for using one or another one techniques is make. (Author)

  11. Effects of burnup on fuel failure. Power burst tests on fuel rods with 13,000 and 32,000 MWd/MTU burnup

    International Nuclear Information System (INIS)

    Results are presented from preliminary tests designed to investigate the behavior of preirradiated fuel rods under reactivity initiated accident (RIA) conditions. The tests were conducted in 1970 as part of the SPERT/Capsule Driver Core (CDC) program. The report was intended to be published in a series of Idaho Nuclear Corporation Interim Technical Reports (IN-ITRs); however, the CDC program was terminated before the report could be released. In September 1975, the Nuclear Regulatory Commission concluded that the data contained in the report could be a valuable reference in planning future water reactor safety program tests and requested its release

  12. The Japanese utilities' requirements for a next century BWR

    International Nuclear Information System (INIS)

    containment vessel failure probability lower than that of ABWR by adopting accident management, use active safety systems for the short-term cooling and a passive system as a backup for active systems for long-term cooling under accident conditions. Core design; A key emphasis for the core design is flexibility in order to address changeable social conditions from the following aspects: flexibility in fuel discharge burnup, flexibility in use of UO2 and MOx fuel, flexibility in operating cycle length, reduced fuel shuffling period to shorten the annual inspection, simplified reactor internals and control rod drive in order to reduce maintenance and inspection work loads. (author)

  13. Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

    1998-04-01

    Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

  14. Processing of the GALILEO fuel rod code model uncertainties within the AREVA LWR realistic thermal-mechanical analysis methodology

    International Nuclear Information System (INIS)

    The availability of reliable tools and associated methodology able to accurately predict the LWR fuel behavior in all conditions is of great importance for safe and economic fuel usage. For that purpose, AREVA has developed its new global fuel rod performance code GALILEO along with its associated realistic thermal-mechanical analysis methodology. This realistic methodology is based on a Monte Carlo type random sampling of all relevant input variables. After having outlined the AREVA realistic methodology, this paper will be focused on the GALILEO code benchmarking process, on its extended experimental database and on the GALILEO model uncertainties assessment. The propagation of these model uncertainties through the AREVA realistic methodology is also presented. This GALILEO model uncertainties processing is of the utmost importance for accurate fuel design margin evaluation as illustrated on some application examples. With the submittal of Topical Report GALILEO to the U.S. NRC in 2013, GALILEO and its methodology are on the way to be industrially used in a wide range of irradiation conditions. (authors)

  15. Characterization of spent fuel disassembly hardware and nonfuel bearing components and their relationship to 10 CFR 61

    International Nuclear Information System (INIS)

    There are a variety of wastes that will be disposed of by the federal waste management system under the Nuclear Waste Policy Act of 1982. The primary waste form is spent nuclear fuel. Currently, this is in the form of fuel assemblies. If the fuel pins are removed from the fuel assembly, as in consolidation, then the f