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Sample records for fuel rod cutting

  1. Development of cutting device for irradiated fuel rod

    International Nuclear Information System (INIS)

    Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

    2004-01-01

    Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

  2. Development of the down-ender and the spent fuel rod cutting device

    International Nuclear Information System (INIS)

    Kim, S. H.; Yoon, Ji Sup; Kim, Young Hwan; Hoo, Jung Jae; Hong, Dong Hee; Kim, Do Woo

    2000-07-01

    It is necessary to disassemble the spent fuel assembly for the recycling of the PWR spent fuels. The spent fuel disassembling process includes transportation and handling of the spent fuel assembly, extraction and cutting of the spent fuel rods, and extraction of the spent fuel pellets(decladding). In this study, the downender of the spent fuel assembly and the spent fuel rod cutting device have been developed. The downender is used to change the posture of the spent fuel assembly from the vertical to the horizontal directions, prior to extracting the fuel rods. The concepts of the remote operation and maintenance has been introduced in the design of the downender. Also, the several design consideration has been given such as the reliable adaptation of the vertically accessing the assembly to the device, the minimization of the shock force when settling down the assembly, and the interface with the rod extraction device without intermittent operation. The spent fuel rod cutting device using a tube cutter is developed for cutting the fuel rods to the suitable size. In designing this device, the mechanical property of the spent fuel rod is examined such as the strength of the clad material and the optimal size of the rod for the extracting process. Also, several cutting methods, which are commercially available, are investigated and tested in terms of the durability, the deformation on the cutting surface of the rods, and the amount of the generated debris, and the fire risk. As like the downender, the design of this device accommodates the concepts of the remote operation and maintenance

  3. A system automatic study for the spent fuel rod cutting and simulated fuel pellet extraction device

    International Nuclear Information System (INIS)

    Jeong, J. H.; Yun, J. S.; Hong, D. H.; Kim, Y. H.; Park, K. Y.

    2001-01-01

    A fuel pellet extraction device of the spent fuel rods is described. The device consists of a cutting device of the spent fuel rods and the decladding device of the fuel pellets. The cutting device is to cut a spent fuel rod to n optimal size for fast decladding operation. To design the device, the fuel rod properties are investigated including the dimension and material of fuel rod tubes and pellets. Also, various methods of existing cutting method are investigated. The design concepts accommodate remote operability for the Hot-Cell(radioactive ) area operation. Also, the modularization of the device structure is considered for the easy maintenance. The decladding device is to extract the fuel pellet from the rod cut. To design this device, the existing method is investigated including the chemical and mechanical decladding methods. From the view point of fuel recovery and feasibility of implementation. it is concluded that the chemical decladding method is not appropriate due to the mass production of radioactive liquid wastes, in spite of its high fuel recovery characteristics. Hence, in this paper, the mechanical decladding method is adopted and the device is designed so as to be applicable to various lengths of rod-cuts. As like the cutting device,the concepts of remote operability and maintainability is considered. Both devices are fabricated and the performance is investigated through a series of experiments. From the experimental result, the optimal operational condition of the devices is established

  4. Development of the spent fuel rod cutting device using the blade cutters

    International Nuclear Information System (INIS)

    Jung, Jae Hoo; Yoon, Ji Sup; Hong, Dong Hee; Kim, Young Hwan; Park, Gee Yong; Kim, Do Woo

    2000-11-01

    A spent fuel rod cutting device is to cut a spent nuclear fuel rod to optimal size for consequent decladding operation. In this paper, various properties of fuel rod, such as a dimension and material of zircaloy tubes and fuel pellets, are investigated. Also, commercially available cutting method and tools is investigated in terms of its performance. As a result, the blade cutter is selected for the design. In order to fabricate the durable blade cutter, various materials are analyzed in terms of material properties, cutter shape, and heat treatment method, etc. Also, the durability of this tool is tested by cutting the SUS tubes and zircaloy tubes. In the device design, the remote maintainability is considered so that the modularized design is accomplished. Also, the other factors considered in the design are the round shape sustainability at the cut surface, the amount of debris generation, and the fire risk, etc. Considering these design consideration, the spent fuel rod cutting device is fabricated and tested

  5. Cutting system for burnable poison rod

    International Nuclear Information System (INIS)

    Shiina, Atsushi; Toyama, Norihide; Koshino, Yasuo; Fujii, Toshio

    1989-01-01

    Burnable poison rods attached to spent fuels are contained in a containing box and transported to a receiving pool. The burnable poison rod-containing box is provisionally situated by the operation to a handling device to a provisional setting rack in a cutting pool and attached to a cutting guide of a cutting device upon cutting. The burnable poison rod is cut only in a cutting pool water and tritium generated upon cutting is dissolved into the cutting pool water. Diffusion of tritium is thus restricted. Further, the cutting pool is isolated by a partition device from the receiving pool during cutting of the burnable poison rod. Accordingly, water in which tritium is dissolved is inhibited from moving to the receiving pool and prevail of tritium contamination can be avoided. (T.M.)

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

    International Nuclear Information System (INIS)

    Lee, Jong Youl; Jung, Jae Hoo; Hong, Dong Hee; Yoon, Ji Sup; Lee, Eun Pyo

    2001-03-01

    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

  7. Cutting method and cutting device for spent fuel rod of nuclear reactor

    International Nuclear Information System (INIS)

    Komatsu, Masahiko; Ose, Toshihiko.

    1996-01-01

    A control rod transferred under water in a vertically suspended state is postured horizontally at such a water depth that radiations can be shielded, and then it is cut to a dropping speed limiting portion and a cross-like main body. The separated cross-like main body portion is further cut in the longitudinal direction and separated into a pair of cut pieces each having an L-shaped cross section. A disk like metal saw is used as a cutting tool. Alternatively, a plasma jet cutter or a melting-type water jet cutter is used as a cutting tool. Then, since the spent control rod to be cut is postured horizontally under water, the water depth for the cutting position can be reduced. As a result, the cutting state using the cutting tool can be observed by naked eyes from the position above the water surface thereby enabling to perform the cutting operation reliably. (N.H.)

  8. Method of inserting fuel rod

    International Nuclear Information System (INIS)

    Kamimoto, Shuji; Imoo, Makoto; Tsuchida, Kenji.

    1991-01-01

    The present invention concerns a method of inserting a fuel rod upon automatic assembling, automatic dismantling and reassembling of a fuel assembly in a light water moderated reactor, as well as a device and components used therefor. That is, a fuel rod is inserted reliably to an aimed point of insertion by surrounding the periphery of the fuel rod to be inserted with guide rods, and thereby suppressing the movement of the fuel rod during insertion. Alternatively, a fuel rod is inserted reliably to a point of insertion by inserting guide rods at the periphery of the point of insertion for the fuel rod to be inserted thereby surrounding the point of insertion with the guide rods or fuel rods. By utilizing fuel rods already present in the fuel assembly as the guide rods described above, the fuel rod can be inserted reliably to the point of insertion with no additional devices. Dummy fuel rods are previously inserted in a fuel assembly which are then utilized as the above-mentioned guide rods to accurately insert the fuel rod to the point of insertion. (I.S.)

  9. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  10. Development on the High-throughput Vol-oxidizer for Decladding and Voloxidation of Spent Fuel Rod-cuts

    International Nuclear Information System (INIS)

    Kim, Young Hwang; Jung, Jae Hoo; Kim, Ki Ho; Park, Byung Buk; Lee, Hyo Jik; Kim, Sung Hyun; Park, Hee Sung; Lee, Jong Kwang; Kim, Ho Dong

    2009-12-01

    A high-throughput vol-oxidizer which can handle a several ten kg HM/batch is being developed to supply U 3 O 8 powders to an electrolytic reduction reactor in pyro-processing. At the first year step(2007), for enhancement of oxidation and recovery rate, we analyzed the mechanical and chemical methods, and devised the main mechanism with ball drop methods and rotary kiln type. Also, the main devices for oxidation and recovery of rod-cuts were designed by using the Solid Works and COSMOS program tools, and manufactured after thermal/mechanical analysis. In order to verify the main devices, simulation fuels(W 90%+SiO 2 10%) were manufactured and the main devices were tested for the oxidation and recovery rate of its. Here the expansion ratio of simulation fuel is similar to U 3 O 8 (2.7). At the second year step(2008), with the constant ration of rod-cuts volume and expansion ratio of U 3 O 8 (2.7), we produced a theoretical equation that can estimate the volume of rod-cuts according to a variation of their weight and lengths. We considered various materials such as ceramics and Ni-Cr, finally, the APM material which can constantly maintain against high temperature(1,200 .deg. C) and vacuum(1 torr) was selected and a vol-oxidizer was designed. At the third year step(2009), in order to manufacture a high-throughput vol-oxidizer, we have analyzed the vol-oxidizer for remote operability and maintainability, also the remote assembling and disassembling possibilities of the selected modules have been analyzed in terms of visibility, interference, approach, weight, and so on. We have presented final modular design and manufactured a high-throughput vol-oxidizer. Also, we have conducted the blank, heating(over 500 .deg. C) and hull separation test(capacity : 50 kg HM/batch, hull length 50mm) on the high-throughput vol-oxidizer. Also, these design technologies for the high-throughput vol-oxidizer will be utilized in the development of a more efficient vol-oxidizer with higher

  11. Failed fuel rod detector

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Katsuya; Matsuda, Yasuhiko

    1984-05-02

    The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

  12. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.; Macivergan, R.; Mckenzie, G.W.

    1980-01-01

    An apparatus incorporating a microprocessor control is provided for automatically loading nuclear fuel pellets into fuel rods commonly used in nuclear reactor cores. The apparatus comprises a split ''v'' trough for assembling segments of fuel pellets in rows and a shuttle to receive the fuel pellets from the split ''v'' trough when the two sides of the split ''v'' trough are opened. The pellets are weighed while in the shuttle, and the shuttle then moves the pellets into alignment with a fuel rod. A guide bushing is provided to assist the transfer of the pellets into the fuel rod. A rod carousel which holds a plurality of fuel rods presents the proper rod to the guide bushing at the appropriate stage in the loading sequence. The bushing advances to engage the fuel rod, and the shuttle advances to engage the guide bushing. The pellets are then loaded into the fuel rod by a motor operated push rod. The guide bushing includes a photocell utilized in conjunction with the push rod to measure the length of the row of fuel pellets inserted in the fuel rod

  13. Fuel rod leak detector

    International Nuclear Information System (INIS)

    Womack, R.E.

    1978-01-01

    A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

  14. Repairs on underwater spent fuel transfer buggy and review of other underwater facilities of Cirus rod cutting building

    International Nuclear Information System (INIS)

    Rao, D.V.H.; Ganeshan, P.; Khadilkar, M.G.

    1994-01-01

    Cirus rod cutting building is a pool of water in concrete unlined bays. This houses several equipment required for processing of spent fuel and other experimental assemblies. These have been in use for over three decades. Recently the fuel transfer buggy had a major breakdown and the repair involved elaborate planning preparation and special methods to ensure safe working condition and to minimise manrem consumption. This also provided an opportunity to assess the condition of other underwater components in radiation environment which were hitherto inaccessible. This paper highlights the repair work carried on buggy and also the effect of ageing on some of the equipment vis a vis the possibility of their life extension. (author). 7 figs

  15. Apparatus for loading fuel pellets in fuel rods

    International Nuclear Information System (INIS)

    Tedesco, R.J.

    1976-01-01

    An apparatus is disclosed for loading fuel pellets into fuel rods for a nuclear reactor including a base supporting a table having grooves therein for holding a multiplicity of pellets. Multiple fuel rods are placed in alignment with grooves in the pellet table and a guide member channels pellets from the table into the corresponding fuel rods. To effect movement of pellets inside the fuel rods without jamming, a number of electromechanical devices mounted on the base have arms connected to the lower surface of the fuel rod table which cyclically imparts a reciprocating arc motion to the table for moving the fuel pellets longitudinally of and inside the fuel rods. These electromechanical devices include a solenoid having a plunger therein connected to a leaf type spring, the arrangement being such that upon energization of the solenoid coil, the leaf spring moves the fuel rod table rearwardly and downwardly, and upon deenergization of the coil, the spring imparts an upward-forward movement to the table which results in physical displacement of fuel pellets in the fuel rods clamped to the table surface. 8 claims, 6 drawing figures

  16. Fuel rod technology

    International Nuclear Information System (INIS)

    Bezold, H.; Romeiser, H.J.

    1979-07-01

    By extensive mechanization and automation of the fuel rod production, also at increasing production numbers, an efficient production shall be secured, simultaneously corresponding to the high quality standard of the fuel rods. The works done up to now concentrated on the lay out of a rough concept for a mechanized production course. Detail-studies were made for the problems of fuel rod humidity, filling and resistance welding. Further promotion of this project and thus further report will be stopped, since the main point of these works is the production technique. (orig.) [de

  17. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    International Nuclear Information System (INIS)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland; Helmut Kuhl

    2015-01-01

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs

  18. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

    2015-05-15

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

  19. Cutting of Control Rod Guide Tubes and Jet Pumps at Wuergassen NPP/ Germany

    International Nuclear Information System (INIS)

    Ralf Loock

    2006-01-01

    KWW Nuclear Power Plant has been erected and put into operation in the time period 1968 to end of 1971. In 23 years of operation about 73 billions kWh electric power have been generated. The electric rated power was 670 MW. In May 1995 the decommission was started for economical reasons, since April 1997 the demolition is in progress. In autumn 2003 FANP has been called upon to cut the Control Rod Guide Tubes and Jet Pumps. After a 10 months preparatory phase, cutting of the Control Rod Guide Tubes and Jet Pumps was started in September 2004. For radiation protection reasons the Control Rod Guide Tubes and Jet Pumps were cut and packed remote controlled under water. The cutting process had been selected particularly under consideration of radiation protection aspects. A combination of the technologies band sawing and nibbling was assigned for cutting of the Control Rod Guide Tubes and Jet Pumps. The band saw consists of a substructure and of a feed system installed on it. The feed system can be equipped with one or two band saws optionally. A turntable is integrated in the substructure. It is used to support the CRGT respectively the residual piece of the CRGT and to move it out of the cut position after finishing the cutting. After getting the agreement to the specification, the equipment has been designed and manufactured according to the specification preferences. After finishing the internal testing, the factory acceptance test as well as acceptance and functional tests of every facility were carried out. After cold testing being performed successfully, the cutting equipment was installed in the NPP and checked for operational safety additionally. Cutting of the Control Rod Guide Tubes and Jet Pumps has been carried out on two working places inside the fuel element storage pool. During the total project duration on site the processing has been supervised by the radiation protection staff permanently. As a result 110 Control Rod Guide Tubes and 18 Jet Pumps were

  20. Inspecting method for fuel rods

    International Nuclear Information System (INIS)

    Watanabe, Masaaki; Kogure, Sumio.

    1976-01-01

    Purpose: To precisely detect the response of flaw in clad tube and submerged fuel pellets from a relationship between the surface of fuel rod and internal signal. Constitution: Ultrasonic reflected waves from the surface of fuel rods and the interior are detected and either one of fuel rod or ultrasonic flaw detecting contact is rotated to thereby precisely detect the response of the flaw of clad tube and submerged fuel pellets from a relationship between said surface and the interior. It will be noted that the ultrasonic flaw detecting contact used is of the line-focus type, the incident angle of ultrasonic wave from the ultrasonic flaw detecting contact relative to the fuel rod is the angle of skew, that is, the ultrasonic flaw detecting contact is not perpendicular to a center axis of the fuel rod but is slightly displace. That is, the use of the aforesaid contact may facilitate discrimination between the surface flaw of the fuel rod and the response of submergence, and in addition, the employment of the aforesaid incident angle makes it hard to receive reflected waves from the surface of the fuel rod which is great in terms of energy to facilitate discrimination of waves responsive to submergence. (Kawakami, Y.)

  1. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

    Lambert, D.V.; Nylund, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

    1991-01-01

    This patent describes a method for loading fuel rods in a desired pattern. It comprises providing a supply of fuel rods of known enrichments; providing a magazine defining a matrix of elongated slots open at their forward ends for receiving fuel rods; defining a fuel rod feed path; receiving successively one at a time along the feed path fuel rods selected from the supply thereof; verifying successively one at a time along the feed path the identity of the selected fuel rods, the verifying including blocking passage of each selected fuel rod along the feed path until the identity of each selected fuel rod is confirmed as correct; feeding to the magazine successively one at a time along the feed path the selective and verified fuel rods; and supporting and moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

  2. Method for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    In a nuclear reactor system which requires periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described. The method consists of: (1) removing the top end from the fuel rod assembly; (2) passing each of multiple fuel rod pulling elements in sequence through a fuel rod container and thence through respective consolidating passages in a fuel rod directing chamber; (3) engaging one of the pulling elements to the top end of each of the fuel rods; (4) drawing each of the pulling elements axially to draw the respective engaged fuel rods in one axial direction through the respective the passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another in the one axial direction into the fuel rod container while maintaining the compacted configuration whereby the fuel rods are aligned within the container in a fuel rod density of the the fuel rod assembly

  3. Inspection system for Zircaloy clad fuel rods

    International Nuclear Information System (INIS)

    Yancey, M.E.; Porter, E.H.; Hansen, H.R.

    1975-10-01

    A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

  4. Vibrational characteristics and wear of fuel rods

    International Nuclear Information System (INIS)

    Schmugar, K.L.

    1977-01-01

    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. Preliminary nuclear design for test MOX Fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

    1997-10-01

    As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

  6. Simulation of leaking fuel rods

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

    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 123 I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123 I release from failed fuel rods during transients

  7. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

    2003-01-01

    The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

  8. Photonic mesophases from cut rod rotators

    Energy Technology Data Exchange (ETDEWEB)

    Stelson, Angela C.; Liddell Watson, Chekesha M., E-mail: cml66@cornell.edu [Materials Science and Engineering, Cornell University, Ithaca, New York 14853 (United States); Avendano, Carlos [Chemical Engineering and Analytical Science, The University of Manchester, Manchester M13 9PL (United Kingdom)

    2016-01-14

    The photonic band properties of random rotator mesophases are calculated using supercell methods applied to cut rods on a hexagonal lattice. Inspired by the thermodynamic mesophase for anisotropic building blocks, we vary the shape factor of cut fraction for the randomly oriented basis. We find large, stable bandgaps with high gap isotropy in the inverted and direct structures as a function of cut fraction, dielectric contrast, and filling fraction. Bandgap sizes up to 34.5% are maximized at high dielectric contrast for rods separated in a matrix. The bandgaps open at dielectric contrasts as low as 2.0 for the transverse magnetic polarization and 2.25 for the transverse electric polarization. Additionally, the type of scattering that promotes the bandgap is correlated with the effect of disorder on bandgap size. Slow light properties are investigated in waveguide geometry and slowdown factors up to 5 × 10{sup 4} are found.

  9. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  10. Spacers for fuel rod clusters

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    The proposition deals with the fixing of nuclear fuel element rods in a grid which consists of a number of crossed Zy-plates which form cells. The rectangular cells have projections which serve as spacers for the fuel rods. According to the invention there are additional butt straps which can be moved in such a way that insertion and extraction of the fuel rods can be done without obstruction and they can be spring-loaded hold in their final position. (UWI) [de

  11. Fabrication of internally instrumented reactor fuel rods

    International Nuclear Information System (INIS)

    Schmutz, J.D.; Meservey, R.H.

    1975-01-01

    Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

  12. Segmented fuel and moderator rod

    International Nuclear Information System (INIS)

    Doshi, P.K.

    1987-01-01

    This patent describes a continuous segmented fuel and moderator rod for use with a water cooled and moderated nuclear fuel assembly. The rod comprises: a lower fuel region containing a column of nuclear fuel; a moderator region, disposed axially above the fuel region. The moderator region has means for admitting and passing the water moderator therethrough for moderating an upper portion of the nuclear fuel assembly. The moderator region is separated from the fuel region by a water tight separator

  13. Detection of failed fuel rods in shrouded BWR fuel assemblies

    International Nuclear Information System (INIS)

    Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

    1988-01-01

    A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

  14. Fuel rod simulator effects in flooding experiments single rod tests

    International Nuclear Information System (INIS)

    Nishida, M.

    1984-09-01

    The influence of a gas filled gap between cladding and pellet on the quenching behavior of a PWR fuel rod during the reflood phase of a LOCA has been investigated. Flooding experiments were conducted with a short length electrically heated single fuel rod simulator surrounded by glass housing. The gap of 0.05 mm width between the Zircaloy cladding and the internal Al 2 O 3 pellets of the rod was filled either wit helium or with argon to vary the radial heat resistance across the gap. This report presents some typical data and an evaluation of the reflood behavior of the fuel rod simulator used. The results show that the quench front propagates faster for increasing heat resistance in the gap between cladding and heat source of the rod. (orig.) [de

  15. Design of active-neutron fuel rod scanner

    International Nuclear Information System (INIS)

    Griffith, G.W.; Menlove, H.O.

    1996-01-01

    An active-neutron fuel rod scanner has been designed for the assay of fissile materials in mixed oxide fuel rods. A 252 Cf source is located at the center of the scanner very near the through hole for the fuel rods. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. We used the Monte Carlo code MCNP to design the scanner and review optimum materials and geometries. An inhomogeneous beryllium, graphite, and polyethylene moderator has been designed that uses source neutrons much more efficiently than assay systems using polyethylene moderators. Layers of borated polyethylene and tungsten are used to shield the detectors. Large NaI(Tl) detectors were selected to measure the delayed gamma rays. The enrichment zones of a thermal reactor fuel pin could be measured to within 1% counting statistics for practical rod speeds. Applications of the rod scanner include accountability of fissile material for safeguards applications, quality control of the fissile content in a fuel rod, and the verification of reactivity potential for mixed oxide fuels. (orig.)

  16. Fuel rod pellet loading head

    International Nuclear Information System (INIS)

    Howell, T.E.

    1975-01-01

    An assembly for loading nuclear fuel pellets into a fuel rod comprising a loading head for feeding pellets into the open end of the rod is described. The pellets rest in a perforated substantially V-shaped seat through which air may be drawn for removal of chips and dust. The rod is held in place in an adjustable notched locator which permits alignment with the pellets

  17. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

    Lambert, D.V.; Nyland, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

    1990-01-01

    This patent describes an apparatus for loading fuel rods in a desired pattern. It comprises: a carousel having a plurality of movable gondolas for stocking thereon fuel rods of known enrichments; an elongated magazine defining a matrix of elongated slots being open at their forward ends for receiving fuel rods; a workstation defining a fuel rod feed path; and a holder and indexing mechanism for movably supporting the magazine and being actuatable for moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

  18. Nuclear reactor fuel rod attachment system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1983-01-01

    The invention involves a technique to quickly, inexpensively and rigidly attach a nuclear reactor fuel rod to a support member. The invention also allows for the repeated non-destructive removal and replacement of the fuel rod. The proposed fuel rod and support member attachment and removal system consists of a locking cap fastened to the fuel rod and a locking strip fastened to the support member or vice versa. The locking cap has two or more opposing fingers shaped to form a socket. The fingers spring back when moved apart and released. The locking strip has an extension shaped to rigidly attach to the socket's body portion

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

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1987-05-01

    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

  20. System for manipulating radioactive fuel rods within a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Tolino, R.W.; King, W.E.; Blickenderfer, J.L.; Roth, C.H. Jr.

    1987-01-01

    A tool is described for manipulating the peripherally located fuel rods of a fuel assembly so that the rods can be visually inspected. The fuel assembly includes top and bottom nozzles, each of which is connected to a support skeleton, as well as grids, and wherein the rods are retained within the grids and confined between the top and bottom nozzles thereof. It consists of: (a) a fixture that is detachably connectable to one of the nozzles of the fuel assembly. The fixture having holes therein, (b) rotating means pivotally mountable within the holes of the fixture for selectively gripping and rotating the rod, and (c) a displacing means mounted on the fixture for reciprocably displacing the rods within the fuel assembly, including a lifting assembly and a push-down assembly for lifting and pushing down a selected one of the rods, respectively, whereby the rods can be selectively rotated, lifted, and pushed down in order to expose portions of the rods which are normally hidden to visual inspection while the nozzles stay connected to the support skeleton and the rods stay confined between the top and bottom nozzles of the fuel assembly

  1. Refabricated and instrumented fuel rods

    International Nuclear Information System (INIS)

    Silberstein, K.

    2005-01-01

    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)

  2. Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-10-01

    Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

  3. Microcomputer system for controlling fuel rod length

    International Nuclear Information System (INIS)

    Meyer, E.R.; Bouldin, D.W.; Bolfing, B.J.

    1979-01-01

    A system is being developed at the Oak Ridge National Laboratory (ORNL) to automatically measure and control the length of fuel rods for use in a high temperature gas-cooled reactor (HTGR). The system utilizes an LSI-11 microcomputer for monitoring fuel rod length and for adjusting the primary factor affecting length. Preliminary results indicate that the automated system can maintain fuel rod length within the specified limits of 1.940 +- 0.040 in. This system provides quality control documentation and eliminates the dependence of the current fuel rod molding process on manual length control. In addition, the microcomputer system is compatible with planned efforts to extend control to fuel rod fissile and fertile material contents

  4. Failure position detection device for nuclear fuel rod

    International Nuclear Information System (INIS)

    Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

    1987-01-01

    Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.)

  5. Failure position detection device for nuclear fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

    1987-03-24

    Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.).

  6. Fuel rod design by statistical methods for MOX fuel

    International Nuclear Information System (INIS)

    Heins, L.; Landskron, H.

    2000-01-01

    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, UO 2 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)

  7. Expandable device for a nuclear fuel rod

    International Nuclear Information System (INIS)

    Gesinski, L.T.

    1978-01-01

    A nuclear fuel rod and a device for use within the rod cladding to maintain the axial position of the fuel pellets stacked one atop another within the cladding are described. The device is initially of a smaller external cross-section than the fuel rod cladding internal cross-section so as to accommodate loading into the rod at preselected locations. During power operation the device responds to a rise in temperature, so as to permanently maintain its position and restrain any axial motion of the fuel pellets

  8. Method and apparatus for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    In a nuclear reactor system requiring periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described comprising the steps of: (1) removing the top end from pulling members having electrodes of weld elements in leading ends thereof in sequence through a fuel rod container and thence through respective consolidating passages in a fuel-rod directing chamber; (3) welding the weld elements of the pulling members to the top end of respective fuel rods corresponding to the respective pulling members; (4) drawing each of the pulling members axially to draw the respective engaged fuel rods in one axial direction through the respective passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another to the one axial direction into the fuel rod container while maintaining the compacting configuration in a fuel rod density which is greater than that of the fuel rod density of the fuel rod assembly

  9. Nondestructive assay of HTGR fuel rods

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1974-01-01

    Performance characteristics of three different radioactive source NDA systems are compared for the assay of HTGR fuel rods and stacks of rods. These systems include the fast neutron Sb-Be assay system, the 252 Cf ''Shuffler,'' and the thermal neutron PAPAS assay system. Studies have been made to determinethe perturbation on the measurements from particle size, kernel Th/U ratio, thorium content, and hydrogen content. In addition to the total 235 U determination, the pellet-to-pellet or rod-to-rod uniformity of HTGR fuel rod stacks has been measured by counting the delayed gamma rays with a NaI through-hole in the PAPAS system. These measurements showed that rod substitutions can be detected easily in a fuel stack, and that detailed information is available on the loading variations in a uniform stack. Using a 1.0 mg 252 Cf source, assay rates of 2 to 4 rods/s are possible, thus facilitating measurement of 100 percent of a plant's throughput. (U.S.)

  10. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Wada, Toyoji.

    1979-01-01

    Purpose: To remove failures caused from combination of fuel-cladding interactions, hydrogen absorptions, stress corrosions or the likes by setting the quantity ratio of uranium or uranium and plutonium relative to oxygen to a specific range in fuel pellets and forming a specific size of a through hole at the center of the pellets. Constitution: In a fuel rods of a structure wherein fuel pellets prepared by compacting and sintering uranium dioxide, or oxide mixture consisting of oxides of plutonium and uranium are sealed with a zirconium metal can, the ratio of uranium or uranium and plutonium to oxygen is specified as 1 : 2.01 - 1 : 2.05 in the can and a passing hole of a size in the range of 15 - 30% of the outer diameter of the fuel pellet is formed at the center of the pellet. This increases the oxygen partial pressure in the fuel rod, oxidizes and forms a protection layer on the inner surface of the can to control the hydrogen absorption and stress corrosion. Locallized stress due to fuel cladding interaction (PCMI) can also be moderated. (Horiuchi, T.)

  11. 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.

  12. Gas pressure and gas purity analyzing device in nuclear fuel rod

    International Nuclear Information System (INIS)

    Mizutani, Chihiro; Hasegawa, Toru.

    1996-01-01

    The present invention provides a device for measuring and analyzing a pressure and a purity of a helium gas sealed in a BWR type nuclear fuel rod. Namely, a portion between a rotational shaft of an electromotive drill for perforating the fuel rod and a vacuum chamber is sealed with a magnetic fluid sealing material so that error factors can be recognized before and after the destruction detection (perforation) of a fuel rod. With such procedures, involving of an atmospheric air from the drill rotational shaft upon perforation can be eliminated. As a result, accuracy for the measurement can be improved. In addition, a filter is disposed to a pipeline connecting the vacuum chamber and the measuring system. With such a constitution, scattering of cutting dusts to the measuring system, troubles due to damages of a stop valve can be reduced. As a result, the efficiency of the measurement is improved. Further, a plurality kinds of gas collecting vessel having different capacities are connected in parallel to the pipeline of the measuring system. Then, the gas collecting vessels can be used selectively. As a result, the device can cope with a gas pressure over a wide range. (I.S.)

  13. Relation of fuel rod service parameters and design requirements to produced fuel rod and their components

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.

    1999-01-01

    Based on the presented material it is possible to state that there is a very close link between the fuel operational parameters and the requirements for its design and production process. The required performance and life-time of a fuel rod can be only assured by the correctly selected design and process solutions. The economical aspect of this problem is significantly depend on the commercial feasibility of a particular selected solution with the provision of an automated and comparative by inexpensive production of a fuel rod and its components. The operational conditions are also important for the life time of the fuel rods. If there are no special measures for the mitigation of the certain operation conditions the leakage of fuel elements can take place before the planned time. (authors)

  14. Apparatus for inspecting a irradiated nuclear fuel rod

    International Nuclear Information System (INIS)

    Saura, Hideaki; Yonemura, Eizo.

    1975-01-01

    Object: To increase safety and inspection efficiency by operating irradiated fuel rods, which are accommodated in a water-filled pool after being taken out from the reactor. Structure: When making inspection of irradiated fuel rods, particularly the cladding tube thereof, a fuel box which stores irradiated fuel rods in a water pool is secured to a securement mechanism with slime removal apparatus and inspection apparatus on either side capable of being vertically moved, and it is then stopped at a water depth of about 2 meters. When the lid of the box is opened, irradiated fuel rods are taken out with gripping means and then secured together with the gripping means to an operation base provided on the outside of the pool. Thereafter, the box is lowered by operating pedals on the operation base to completely pull out the irradiated fuel rods from the box, and the irradiated fuel rods are then horizontally moved and then held in a suspended state. Next a slime removal apparatus in raised by operating pedals and an inspection element assembly are progressively raised for inspection of the state of the cladding tube of each fuel rod after removal of slime therefrom. (Nakamura, S.)

  15. 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.

  16. Apparatus for loading fuel rods into grids of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Ferlan, S.J.

    1989-01-01

    For use with a nuclear fuel assembly including support grids having cells for receiving fuel rods and with detents disposed within the respective cells for resiliently engaging and laterally supporting the fuel rods received therein, an apparatus is described for facilitating scratchless insertion of each fuel rod into cells of the support rids. The apparatus consists of: a thin-walled metallic tubular member which is long enough to extend through at least a majority of support grids, and is positionable so as to have its thin wall interposed, during insertion of each fuel rod, between the latter and the detents within the cells receiving it, the thin-walled tubular member having a substantially uniform wall thickness of not more than about 0.008 inch, an as-formed inner diameter substantially equal to the outer diameter of the fuel rod, and a longitudinal slit formed in the wall of the tubular member so as to render the wall resiliently deflectable in a diameter-reducing sense, the longitudinal slit having a width sufficient to preclude overlapping of the edges of the wall along the slit, and insufficient for any of the detents to enter the slit when the wall of the tubular member is in position between the detents and the fuel rod

  17. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  18. Process development and fabrication for sphere-pac fuel rods

    International Nuclear Information System (INIS)

    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

  19. Fabrication of preliminary fuel rods for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

    2012-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  20. Processing of poison rods with a view to disposal

    International Nuclear Information System (INIS)

    Bichet, R.; Charamathieu, A.; Lasseur, C.; Golicheff, I.; Pouteaux, M.

    1979-01-01

    In the core of the French 900 and 1300 MW reactors, a certain number of rods have to be processed as wastes, particularly the burnable poison rods used during reactor start-up (900 MW: 68 rods; 1300 MW: 96 rods). Several solutions are possible: cutting and conditionning in reactor pool; transfer of the poison rods to a cutting and conditionning facility; transfer of the poison rods and fuel assemblies to a storage area where they are cutted and stored. Each of these solutions are studied, the advantages and disadvantages being presented

  1. 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.

  2. 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.

  3. LOFT advanced fuel rod instrumentation development

    International Nuclear Information System (INIS)

    Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

    1978-01-01

    Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

  4. Engineering design on main mechanism of a high throughput vol-oxidizer for decladding and vol-oxidation of rod-cuts

    International Nuclear Information System (INIS)

    Kim, Y. H.; Park, B. S.; Jung, J. H.; Yoon, J. S.; Kim, H. D.; Hwang, J. S.; Yoon, K. H.

    2008-01-01

    In this paper, we designed the main mechanisms for a high throughput device for the rod-cuts of a spent fuel. To design the main mechanisms, we evaluated the current mechanical (slitting, ball mill, roller straightening) and chemical methods (muffle furnace, rotary kiln). As a result, the methods for a ball drop and a rotary drum as concepts were selected at the analysis step. For an enhancement of the oxidation rate, we devised blades for the reactor as a mesh type. Also, for an enhancement of the decladding rate, we designed the ball size and the rotation of the reactor as a mesh type and devised a vacuum system for the fission products. We also designed the main mechanisms devices and tested the capacity of these devices. Mechanisms for the oxidation and recovery can simultaneously handle the rod-cuts of a spent fuel and provide an independent recovery. The results of the mechanisms designs can be used for a scale-up of a high throughput device

  5. Fuel followed control rod installation at AFRRI

    International Nuclear Information System (INIS)

    Moore, Mark; Owens, Chris; Forsbacka, Matt

    1992-01-01

    Fuel Followed Control Rods (FFCRs) were installed at the Armed Forces Radiobiology Research Institute's 1 MW TRIGA Reactor. The procedures for obtaining, shipping, and installing the FFCRs is described. As part of the FFCR installation, the transient rod drive was relocated. Core performance due to the addition of the fuel followed control rods is discussed. (author)

  6. Inlet for fuel assembly having finger control rods

    International Nuclear Information System (INIS)

    Berglund, A.; Suvanto, A.; Tornblom, L.

    1975-01-01

    A nuclear reactor with vertically arranged fuel assemblies positioned on supporting members and with control rods displaceably arranged in guide tubes between the fuel rods inside the fuel assemblies is described. The supporting plate is provided with a transverse end piece with throttling means for the liquid flow which passes from below up through the supporting member and past the fuel rods in the fuel assembly. The inlets for the guide tubes for the control rods are located below the end piece and the throttling means. In this way a higher pressure prevails at the inlet to the guide tubes than above the end piece, so that a stronger flow of coolant is produced through guide tubes than through the fuel assembly. (U.S.)

  7. BWR fuel assembly having fuel rod spacers axially positioned by exterior springs

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1988-01-01

    In a fuel assembly having spaced fuel rods, an outer hollow tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid there-along, and at least one spacer being disposed along the channel and about the fuel rods so as to maintain them in side-by-side spaced relationship, an arrangement for disposing the spacer in a desired axial position along the fuel rods is described comprising: yieldably resilient springs disposed between an interior side of the outer channel and an exterior side of the spacer. The springs have an inherent spring bias directed away from the exterior sides of the spacers and toward the interior side of the channel such that by contact with the channel and spacer the springs assume states in which they are deflected away from the channel interior side so as to exert sufficient compressive contacting force thereon to maintain the spacer substantially stationary in the desired axial position along the fuel rods

  8. 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.

  9. Experiment studies of fuel rod vibration in coolant flow for substantiation of vibration stability of fuel rods with no fretting-wear

    International Nuclear Information System (INIS)

    Egorov, Yu. V.; Afanasiev, A. V.; Makarov, V. V.; Matvienko, I. V.

    2013-01-01

    For substantiation of vibration stability it is necessary to determine the ultimate permissible vibration levels which do not cause fretting, to compare them with the level of fuel rod vibration caused by coolant flow. Another approach is feasible if there is experience of successful operation of FA-prototypes. In this case in order to justify vibration stability it may be sufficient to demonstrate that the new element does not cause increased vibration of the fuel rod. It can be done by comparing the levels of hydro-dynamic fuel rod vibration and FA new designs. Program of vibration tests of TVS-2M model included studies of forced oscillations of 12 fuel rods in the coolant flow in the spans containing intensifiers, in the reference span without intensifiers, in the lower spans with assembled ADF and after its disassembly. The experimental results for TVS-2M show that in the spans with intensifier «Sector run» the level of movements is 6% higher on the average than in the span without intensifiers, in the spans with intensifier «Eddy» it is 2% higher. The level of fuel rod vibration movements in the spans with set ADF is 2 % higher on the average than without ADF. During the studies of TVS-KVADRAT fuel rod vibration, the following tasks were solved: determination of acceleration of the middle of fuel rod spans at vibration excited due to hydrodynamics; determination of influence of coolant thermal- hydraulic parameters (temperature, flowrate, dynamic pressure) on fuel rod vibration response; determination of influence of span lengths on the vibration level. Conclusions: 1) The vibration tests of the full-scale model of TVS-2M in the coolant flow showed that the new elements of TVS-2M design (intensifiers of heat exchange and ADF) are not the source of fuel rod increased vibration. Considering successful operation of similar fuel rod spans in the existing TVS-2M design, vibration stability of TVS-2M fuel rods with new elements is ensured on the mechanism of

  10. Process and equipment for locating defective fuel rods of a reactor fuel element

    International Nuclear Information System (INIS)

    Jester, A.; Honig, H.

    1977-01-01

    By this equipment, well-known processes for determining defective fuel rods of a reactor fuel element are improved in such a fashion that defective fuel rods can be located individually, so that it is possible to replace them. The equipment consists of a cylindrical test vessel open above, which accommodates the element to be tested, so that an annular space is left between the latter's external circumference and the wall of the vessel, and so that the fuel rods project above the vessel. A bell in the shape of a frustrum of a cone is inverted over the test vessel, which has an infra-red measuring equipment at a certain distance above the tops of the fuel rods. The fuel element to be tested together with the test vessel and hood are immersed in a basin full of water, which displaces water by means of gas from the hood. The post-shutdown heat increases the temperature in the water space of the test vessel, which is stabilised at 100 0 C. In each defective fuel rod the water which has penetrated the defective fuel rod previously, or does so now, starts to boil. The steam rising in the fuel rod raises the temperature of the defective fuel rod compared to all the sound ones. The subsequent measurement easily determines this. Where one can expect interference with the measurement by appreciable amounts of gamma rays, the measuring equipment is removed from the path of radiation by mirror deflection in a suitably shaped measuring hood. (FW) [de

  11. Modelling of pellet-cladding interaction for PWRs reactors fuel rods

    International Nuclear Information System (INIS)

    Esteves, A.M.

    1991-01-01

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

  12. Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

    2008-09-01

    The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

  13. Device for detecting defective nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Steven, J.

    1976-01-01

    A moisture sensor is provided for a nuclear fuel rod for water-cooled nuclear reactors wherein moisture can be present. The fuel rod has an end cap and a charge of nuclear fuel. The moisture sensor is disposed between the end cap and the charge and serves to detect a leak in the fuel rod. The moisture sensor includes a capsule-like housing having an inner space and having openings through which moisture can pass into the inner space in the event of a leak in the fuel rod. Ferromagnetic material is disposed in the inner space of the housing together with a moisture detector responsive to moisture for altering the diposition of the ferromagnetic material in the inner space. 5 claims, 6 drawing figures

  14. Apparatus and method for loading fuel rods into grids of a fuel assembly

    International Nuclear Information System (INIS)

    De Mario, E.E.; Burman, D.L.; Olson, C.A.; Secker, J.R.

    1987-01-01

    This patent describes a fuel assembly having fuel rods and at least one grid formed of interleaved straps and yieldable springs, the interleaved straps defining hollow cells aligned in rows and columns thereof for receiving the respective fuel rods. A pair of the springs are disposed within each of the cells for engaging and supporting one of the fuel rods when received in the cell. An apparatus is described for facilitating the loading of the fuel rods into the grid of the fuel assembly, comprising: (a) first mean insertable concurrently into the cells of the grid for engaging and moving the springs from respective first positions in which each pair of springs will engage a respective fuel rod when disposed within the grid cell to respective second positions in which each pair of springs is disengaged from the respective fuel rod when disposed within the grid cell; (b) a pair of second means, one of the pair of the second means being insertable concurrently into the rows of the cells of the grid and the other of the pair of second means being insertable concurrently into the column of the cells

  15. Method and apparatus for inspection of nuclear fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1977-01-01

    A method and apparatus are provided for the inspection of nuclear fuel rods to detect defects or failures in such rods. Assemblies of fuel rods are immersed in water and means are provided for causing a change in the relative pressures in the water and within the fuel rod such that fluid is expelled from the rod through any defects that may exist. Means are also provided for thereafter vibrating the rods to cause additional internal fluid or other material that may be trapped in the rod to be expelled. Sensors are provided for detecting the emission of bubbles of fluid or other material from the rod and for locating the position of the defective rod in the assembly. 5 figures

  16. Rotation of intramedullary alignment rods affects distal femoral cutting plane in total knee arthroplasty.

    Science.gov (United States)

    Maderbacher, Günther; Matussek, Jan; Keshmiri, Armin; Greimel, Felix; Baier, Clemens; Grifka, Joachim; Maderbacher, Hermann

    2018-02-17

    Intramedullary rods are widely used to align the distal femoral cut in total knee arthroplasty. We hypothesised that both coronal (varus/valgus) and sagittal (extension/flexion) cutting plane are affected by rotational changes of intramedullary femoral alignment guides. Distal femoral cuts using intramedullary alignment rods were simulated by means of a computer-aided engineering software in 4°, 6°, 8°, 10°, and 12° of valgus in relation to the femoral anatomical axis and 4° extension, neutral, as well as 4°, 8°, and 12° of flexion in relation to the femoral mechanical axis. This reflects the different angles between anatomical and mechanical axis in coronal and sagittal planes. To assess the influence of rotation of the alignment guide on the effective distal femoral cutting plane, all combinations were simulated with the rod gradually aligned from 40° of external to 40° of internal rotation. Rotational changes of the distal femoral alignment guides affect both the coronal and sagittal cutting planes. When alignment rods are intruded neutrally with regards to sagittal alignment, external rotation causes flexion, while internal rotation causes extension of the sagittal cutting plane. Simultaneously the coronal effect (valgus) decreases resulting in an increased varus of the cutting plane. However, when alignment rods are intruded in extension or flexion partly contradictory effects are observed. Generally the effect increases with the degree of valgus preset, rotation and flexion. As incorrect rotation of intramedullary alignment guides for distal femoral cuts causes significant cutting errors, exact rotational alignment is crucial. Coronal cutting errors in the distal femoral plane might result in overall leg malalignment, asymmetric extension gaps and subsequent sagittal cutting errors.

  17. The effect of the fuel rod friction force to the fuel assembly lateral mechanical characteristics

    International Nuclear Information System (INIS)

    Ha, Dong Geun; Jeon, Sang Youn; Suh, Jung Min

    2012-01-01

    The Fuel Assembly (FA) for light water reactor consists of hundreds of fuel rods, guide tubes, spacer grids, top/bottom nozzles. The guide tubes transmit vertical loads between the top and bottom nozzles, position the fuel rod support grids vertically, react the loads from the fuel rods that are applied to the grids, and provide some of the lateral load capability for the overall fuel assembly. The guide tubes are the structural members of the skeleton assembly. And the spacer grids maintain the fuel rod array by providing positive lateral restraint to the fuel rod but only frictional restraint in the axial direction. Figure 1 shows the outline of skeleton, FA and the location of guide tubes in the view of cross section. 17x17 FA has 24 guide tubes and one instrumentation tube. When the FA is in reactor, the lateral stiffness is one of very important factors from the view point of in reactor integrity of fuel assembly such as guarantee of the cool able geometry, the control rod insertion etc. The lateral stiffness of FA is mainly determined by skeleton lateral stiffness. And the fuel rods loaded in the spacer grids reinforce the FA lateral stiffness. Generally, fuel rods and spacer grids create the nonlinear friction force between fuel rod tube and grid spring/dimple against external lateral force of FA. Thus, it is necessary to study the contribution of the fuel rods friction force to the FA lateral stiffness. So, this paper is to show how much amount of the fuel rod grid interaction contributes to the FA lateral stiffness based on the test results

  18. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

    A nuclear fuel loading apparatus, incorporating a microprocessor control unit, is described which automatically loads nuclear fuel pellets into dual fuel rods with a minimum of manual involvement and in a manner and sequence to ensure quality control and accuracy. (U.K.)

  19. Experimental studies of resistance fretting-wear of fuel rods for VVER-1000 and TVS-KVADRAT fuel assemblies

    International Nuclear Information System (INIS)

    Makarov, V.; Afanasiev, A.; Egorov, Yu.; Matvienko, I.

    2015-01-01

    The paper covers the results of the studies performed to justify the wear resistance of fuel rods in contact with the spacer grids of TVS VVER-1000 fuel assembly and TVS-KVADRAT square fuel assembly of Russian design for PWR-900 reactor. The presented results of three testing stages comprise: Testing of mockup fuel rods of VVER TVS fuel assembly for fretting wear under the conditions of the water chemistry of VVER reactor; Testing models of different design embodiments of the fuel rods for VVER TVS fuel assembly for fretting wear in still cold water; Testing mockup fuel rods of TVS-KVADRAT square fuel assembly for PWR reactor for frettingwear under the conditions of PWR water chemistry. The effect of structural and operational factors was determined (amplitudes, fuel rod vibration frequencies, values of cladding-to-spacer grid cell gap for the depth of fuel rod cladding wear etc.), an assessment was made of the threshold values of fuel rod vibration parameters, which, if not exceeded, provide the absence of the fuel rod cladding fretting wear in the fuel rod-to spacer grid contact area. Key words: fretting wear, fuel rod, spacer grid, VVER, PWR (author)

  20. Fuel rod puncturing and fission gas monitoring system examination techniques

    International Nuclear Information System (INIS)

    Song, Woong Sup

    1999-02-01

    Fission gas products accumulated in irradiated fuel rod is 1-2 cm 3 in CANDU and 40-50 cm 3 in PWR fuel rod. Fuel rod puncturing and fission gas monitoring system can be used for both CANDU and PWR fuel rod. This system comprises puncturing device located at in cell part and monitoring device located at out cell part. The system has computerized 9 modes and can calculate both void volume and mass volume only single puncturing. This report describes techniques and procedure for operating fuel rod puncturing and gas monitoring system which can be play an important role in successful operation of the devices. Results obtained from the analysis can give more influence over design for fuel rods. (Author). 6 refs., 9 figs

  1. SFAK, Unscattered Gamma Self-Absorption from Regular Fuel Rod Assemblies

    International Nuclear Information System (INIS)

    Wand, H.

    1982-01-01

    1 - Description of problem or function: Calculation of the self- absorption of unscattered (gamma-) radiation from fuel assemblies which contain a regular arrangement of identical fuel rods. 2 - Method of solution: The point-kernel is integrated over the radiation sources, i.e. the fuel rods. A uniform mesh of integration points is used for each of the fuel rods. 3 - Restrictions on the complexity of the problem: Number of fuel rods is dynamically allocated

  2. On-line fuel and control rod integrity management in BWRs

    International Nuclear Information System (INIS)

    Larsson, Irina; Sihver, Lembit

    2011-01-01

    Surveillance of fuel and control rod integrity in a BWR core is essential to maintain a safe and reliable operation of a nuclear power plant. An accurate and prompt way to monitor fuel integrity in a reactor core during reactor operation is by using on-line measurements of the gamma emitting noble gas activities in the off-gas system. The integrity of control rods can be efficiently followed by on-line measurements of the helium (He) concentration in the off-gases. This method also gives information about fuel rod failures since He is used as a fill gas in the fuel rods. To survey fuel and control rod integrity during reactor operation, a system consisting of combined gamma and He on-line measurements in the off-gases should be used. Such a system can detect and follow the behavior of fuel and control rod failures. In addition, it can separate fuel failures from control rod failures since fuel rods contain both He and gamma emitting noble gases, while control rods only contain He. Moreover, the system is able to distinguish primary fuel failures from degradation of already existing ones. In this paper we present a combined system for on-line measurements of He and gamma emitting noble gases in the reactor off-gas system and measuring experiences from different BWRs. (author)

  3. Lumped-parameter fuel rod model for rapid thermal transients

    International Nuclear Information System (INIS)

    Perkins, K.R.; Ramshaw, J.D.

    1975-07-01

    The thermal behavior of fuel rods during simulated accident conditions is extremely sensitive to the heat transfer coefficient which is, in turn, very sensitive to the cladding surface temperature and the fluid conditions. The development of a semianalytical, lumped-parameter fuel rod model which is intended to provide accurate calculations, in a minimum amount of computer time, of the thermal response of fuel rods during a simulated loss-of-coolant accident is described. The results show good agreement with calculations from a comprehensive fuel-rod code (FRAP-T) currently in use at Aerojet Nuclear Company

  4. Gamma-ray spectroscopy on irradiated fuel rods

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac

    2009-01-01

    The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

  5. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2014-01-01

    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

  6. 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.

  7. Evolution of fuel rod support under irradiation impact on the mechanical behaviour of fuel assemblies

    International Nuclear Information System (INIS)

    Billerey, Antoine; Waeckel, Nicolas

    2005-01-01

    New fuel management targets imply to increase fuel assembly discharge burnup. Therefore, the prediction of the mechanical behaviour of the irradiated fuel assembly is essential such as excessive fuel assembly distortion induce incomplete Rod Cluster Control Assembly insertion problems (safety issue) or fuel rod vibration induced wear leading to leaking rods (plant operation problems). Within this framework, one of the most important parameter is the knowledge of the fuel rod support in the grid cell because it directly governs the mechanical behaviour of the fuel assembly and consequently allows to predict the behaviour of irradiated structures in terms of (1) axial and lateral deformation (global behaviour of the assembly) and (2) rod vibration induced wear (local behaviour of the rod). Generally, fuel rod support is provided by a spring-dimple system fixed to the grid. During irradiation, the spring force decreases and a gap between the rod and the spring may occur. This phenomenon is due to (1) stress relieving in the spring and in the dimples, (2) grid growth and (3) reduction of the rod diameter. Two models have been developed to predict the behaviour of the rod in the cell. The first model is dedicated to the evaluation of the spring force relaxation during irradiation. The second one can assess the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (1) the creep laws of the grid materials, (2) the growth law of the grid, (3) the evolution of rod diameter and (4) the design of the fuel rod support. The aim of this paper is to: (1) evaluate the consequences of grid support design modifications on the rod vibration sensitivity in terms of predicted rod to grid maximum gap during irradiation and time in operation with an open rod to grid gap, (2) evaluate, using a linear or non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the overall mechanical

  8. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  9. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

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

    International Nuclear Information System (INIS)

    Takase, Kazuyuki

    1988-03-01

    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 -10 Re 1.43 , 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)

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

    International Nuclear Information System (INIS)

    Jones, W.J.

    1979-01-01

    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

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

    International Nuclear Information System (INIS)

    Malang, S.

    1975-11-01

    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

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

    International Nuclear Information System (INIS)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

    2003-11-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    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.

  15. Method for wrapping a wire round a nuclear fuel rod

    International Nuclear Information System (INIS)

    Nakayasu, Fumio.

    1974-01-01

    Object: To provide a method for winding a wire round a nuclear fuel rod with accurate pitches without imparting any local strain or torsion to the wire. Structure: A wire is fixed on one end of the fuel rod, and the other end of the wire is secured to a universal joint leaving a winding allowance to the fuel rod. The wire is linearly stretched by a predetermined tension through the universal joint so as to provide an angle of development theta corresponding to the desired winding pitch, and then, the fuel rod may be rotated so that the end of the wire on the side of the universal joint is moved towards the fuel rod so as to render the angle of development theta constant in proportion to said rotation of the fuel rod. (Kamimura, M.)

  16. Ultrasonics aids the identification of failed fuel rods

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Over a number of years Brown Boveri Reaktor of West Germany has developed and commercialized an ultrasonic failed fuel rod detection system. Sipping has up to now been the standard technique for failed fuel detection, but sipping can only indicate whether or not an assembly contains defective rods; the BBR system can tell which rod is defective. (author)

  17. Potential impacts of crud deposits on fuel rod behaviour on high powered PWR fuel rods

    International Nuclear Information System (INIS)

    Wilson, W.; Comstock, R.J.

    1999-01-01

    Fuel assemblies operating with significant sub-cooled boiling are subject to deposition of surface deposits commonly referred to as crud. This crud can potentially cause concentration of chemical species within the deposits which can be detrimental to cladding performance in PWRs. In addition, these deposits on the surface of the cladding can result in power anomalies and erroneous reporting of fuel rod oxide thickness which can substantially hamper corrosion and core performance modeling efforts. Data is presented which illustrates the importance of accounting for the presence of crud on fuel cladding surfaces. Several methods used to correct for this phenomenon when collecting and analyzing zirconium alloy field oxide thickness measurements are described. Various observations related to crud characteristics and its impact on fuel rod performance are also addressed. (author)

  18. Nuclear fuel assembly with improved spectral shift-producing rods

    International Nuclear Information System (INIS)

    Ferrari, H.M.

    1987-01-01

    This patent describes a nuclear reactor having fuel assemblies and a moderator-coolant liquid flowing through the fuel assemblies, each fuel assembly including an organized array of nuclear fuel rods wherein the moderator-coolant liquid flows along the fuel rods, at least one improved spectral shift-producing rod disposed among the fuel rods. The spectra shift-producing rod consists of: (a) an elongated hollow hermetically-sealed tubular member; (b) a weakened region formed in a portion of the member, the portion being subject to rupture at a given level of internal pressure; and (c) burnable poison material contained in the member which generates gas in the member as operation of the reactor proceeds normally, the material being soluble in the moderator-coolant liquid when brought into contact therewith; (d) the given level of internal pressure being less than the maximum level of internal pressure normally expected to be generated within the member by the poison material by normal operation of the reactor

  19. Simulation of vibration modes of the fuel rod damaged due to the grid-to-rod fretting wear

    International Nuclear Information System (INIS)

    Kim, Kyu Tae; Kim, Kyeong Koo; Jang, Young Ki; Lee, Kyou Seok

    1997-01-01

    The flow-induced fuel fretting wear observed in some PWRs mainly proceeds in the grid-to-rod contact positions. The grid-to-rod fretting wear in the PWR fuel assembly depends on grid-to-rod gap size, its axial profile and flow-induced vibration. This paper describes the GRIDFORCE program which generates the axially dependent grid-to-rod gap size as a function of burnup. The axially dependent grid-to-rod gap profiles are employed to predict the fuel rod vibration mode shapes by the ANSYS code. With the help of the Paidousis empirical formula, this paper also calculates the fuel rod vibration amplitudes under various supporting conditions, which indicates that the increase of the number of unsupported mid-grids will increase the fuel rod vibration amplitude. On the other hand, the comparison of the predicted vibration mode shapes and the observed mid-grid fretting wear pattern indicates that the 1st and 6th vibration mode shapes under the supporting inactive condition at the mid-grids can simulate the observed mid-grid fretting wear profile. This paper also proposes design guidelines against the grid-to-rod fretting wear. (author). 3 refs., 8 figs

  20. Individual nuclear fuel rod weighing system

    International Nuclear Information System (INIS)

    Fogg, J. L.; Howell, C. A.; Smith, J. H.; Vining, G. E.

    1985-01-01

    An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

  1. 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)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Katanishi, S.

    1987-03-01

    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, UO 2 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) [de

  2. The irradiation performance of austenitic stainless steel clade PWR fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

  3. Nuclear fuel rod end plug weld inspection

    International Nuclear Information System (INIS)

    Parker, M. A.; Patrick, S. S.; Rice, G. F.

    1985-01-01

    Apparatus and method for testing TIG (tungsten inert gas) welds of end plugs on a sealed nuclear reactor fuel rod. An X-ray fluorescent spectrograph testing unit detects tungsten inclusion weld defects in the top end plug's seal weld. Separate ultrasonic weld inspection system testing units test the top end plug's seal and girth welds and test the bottom end plug's girth weld for penetration, porosity and wall thinning defects. The nuclear fuel rod is automatically moved into and out from each testing unit and is automatically transported between the testing units by rod handling devices. A controller supervises the operation of the testing units and the rod handling devices

  4. Investigation of water-logged spent fuel rods under dry storage conditions

    International Nuclear Information System (INIS)

    Kohli, R.; Pasupathi, V.

    1986-09-01

    Tests were conducted to determine the amount of moisture contained in breached, water-logged spent fuel rods and the rate of release. Two well-characterized BWR fuel rods with reactor-induced breaches were tested in a hot cell. These rods contained approximately 6 to 10 g of moisture, most of which was released during heating tests simulating normal cask drying operations. Additional testing with two intentionally defected fuel rods (BWR and PWR) was performed to evaluate the effect of the cladding breach on migration of moisture along the length of the fuel rod. The results showed that the moisture released from reactor-breached spent fuel rods was insufficient to cause degradation of fuel or dry storage system components

  5. Fuel rod attachment system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member or vice versa. The locking cap has two opposing fingers and shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed. In an alternative embodiment, the cap is rigid and the strip is transversely resiliently compressible. (author)

  6. Individual nuclear fuel rod weighing system

    International Nuclear Information System (INIS)

    Fogg, J.L.; Smith, J.H.; Vining, G.E.; Howell, C.A.

    1985-01-01

    An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor is discussed. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

  7. Failed fuel rod detection method by ultrasonic wave

    International Nuclear Information System (INIS)

    Takamatsu, Masatoshi; Muraoka, Shoichi; Ono, Yukio; Yasojima, Yujiro.

    1990-01-01

    Ultrasonic wave signals sent from an ultrasonic receiving element are supplied to an evaluation circuit by way of a gate. A table for gate opening and closing timings at the detecting position in each of the fuel rods in a fuel assembly is stored in a memory. A fuel rod is placed between an ultrasonic transmitting element and the receiving element to determine the positions of the transmitting element and the receiving element by positional sensors. The opening and closing timings at the positions corresponding to the result of the detection are read out from the table, and the gates are opened and closed by the timing. This can introduce the ultrasonic wave signals transmitted through a control rod always to the evaluation circuit passing through the gate. Accordingly, the state of failure of the fuel rod can be detected accurately. (I.N.)

  8. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    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). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  9. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myung Sun; Lee, Jong Po; Ju, Young Sang [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-11-01

    Acoustic resonance scattering from a nuclear fuel rod in water is analyzed. A new model for the background which is attributed to the interference of reflected wave and diffracted wave is found and here named {sup t}he inherent background{sup .} The resonance spectrum of a fuel rod is obtained by subtracting the inherent background from the scattered pressure. And also analyzed are the effect of material damping of cladding tube and pellet on the resonance spectrum of a fuel rod. The propagation characteristics of circumferential waves which cause the resonances of cladding tube is produced and the appropriate resonance modes for the application to the inspection of assembled fuel rods are selected. The resonance modes are experimentally measured for pre- and post-irradiated fuel rods and the validation of the fuel rod inspection using ultrasonic resonance phenomenon is examined. And thin ultrasonic sensors accessible into the narrow interval (about 2-3mm) between assembled fuel rods are designed and manufactured. 14 refs. (Author).

  10. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

  11. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  12. Axial transport of fission gas in LWR fuel rods

    International Nuclear Information System (INIS)

    Kinoshita, M.

    1983-01-01

    With regard to fission gas transportation inside the fuel rod, the following three mechanisms are important: (1) a localized and time dependent fission gas release from UO 2 fuel to pellet/clad gap, (2) the consequent gas pressure difference between the gap and the plenum, and (3) the inter-diffusion of initially filled Helium and released fission gas such as Xenon. Among these three mechanisms, the 2nd mechanism would result in the one dimensional flow through P/C gap in the axial direction, while the 3rd would average the local fission gas concentration difference. In this paper, an attempt was made to develop a computerized model, LINUS (LINear flow and diffusion under Un-Steady condition) describing the above two mechanisms, items (2) and (3). The item (1) is treated as an input. The code was applied to analyse short length experimental fuel rods and long length commercial fuel rods. The calculated time evolution of Xe concentration along the fuel column shows that the dilution rate of Xe in commercial fuel rods is much slower than that in short experimental fuel rods. Some other sensitivity studies, such as the effect of pre-pressurization, are also presented. (author)

  13. Effect analysis of air introduced by pressurization on fuel rod performances

    International Nuclear Information System (INIS)

    Ren Qisen; Liu Tong; Sheng Guofu

    2012-01-01

    In the process of pressurization and seal welding, it is common practice to vacuumize before gas filling for the sake of preventing introducing air and other impurities, which would affect the gas composition inside of the fuel rod. However, vacuumization during pressurization is likely not being required sometimes in order to simplify the fabrication procedure. In the present work, based on the AFA3G fuel rod design with 2 MPa of filling gas, analyses on fuel rod performances were carried out under the condition of pressurization with and without vacuumization, respectively. Furthermore, the effect on hydrogen content in fuel rod was preliminarily discussed. Results indicate that the impacts of air composition introduced by pressurization on fuel rod thermal-mechanical performances, such as internal pressure and fuel center temperature, were extremely slight. The gap conductance varies to some extent as a result of the change of gas composition due to air introduced in fuel rod. The impact of humidity on water content in fuel rod is negligible at a low temperature of around 25℃. However, at higher temperature, it is essential to pay attention on the control of fabrication process, and prevent much moisture entering into the fuel rod and increasing the probability of hydriding failure. (authors)

  14. Model of cooling nuclear fuel rod in the nuclear reactor

    International Nuclear Information System (INIS)

    Lavicka, David; Polansky, Jiri

    2010-01-01

    The following topics are described: Some basic requirements for nuclear fuel rods; The VVER 1000 fuel rod; Classification of the two-phase flow in the vertical tube; Type of heat transfer crisis in the vertical tube; Experimental apparatus; Model of the nuclear fuel rod and spacers; Potential of the experimental apparatus (velocity profile measurement via PIV; thermal flow field measurement by the PLIF method; cooling graph in dependence on the fuel rod temperature; comparison of the hydrodynamic properties with respect to the design features of the spacers). (P.A.)

  15. Fuel rod fixing system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

    This is a reusable system for fixing a nuclear reactor fuel rod to a support. An interlock cap is fixed to the fuel rod and an interlock strip is fixed to the support. The interlock cap has two opposed fingers, which are shaped so that a base is formed with a body part. The interlock strip has an extension, which is shaped so that this is rigidly fixed to the body part of the base. The fingers of the interlock cap are elastic in bending. To fix it, the interlock cap is pushed longitudinally on to the interlock strip, which causes the extension to bend the fingers open in order to engage with the body part of the base. To remove it, the procedure is reversed. (orig.) [de

  16. The thermo-mechanics of the PWR fuel rod

    International Nuclear Information System (INIS)

    Barral, J.C.; Gautier, B.; Chaigne, G.

    1999-01-01

    The fuel rod mechanics is of a great importance in the safety and performance of the reactors. In this domain a meeting has been organized by the SFEN the 18 march 1998 at Paris. With the participation of scientists from CEA, EDF and Framatome, the physics of the fuel rods was presented based on four main aspects. Two first papers dealt with the solicitations of the fuel rod in normal and accidental conditions. The physical phenomena under irradiation were then detailed in the four following talks. Three papers presented the simulation and the codes of the fuel-cladding interactions with the diabolo effect. The last paper was devoted to the experiment feedback and the research programs. (A.L.B.)

  17. PCI/SCC failure behavior of KWU/CE fuel rods

    International Nuclear Information System (INIS)

    Kikuchi, Akira

    1983-10-01

    The Over Ramp (Studsvik Over Ramp-STOR) project is an international power ramping irradiation program for studying PCI/SCC failure behavior of PWR-fuel rods. The project had its activities for about three years (Apr., 1977 - Dec., 1980) as the cooperation works of twelve participants composing nine countries. The present report introduces the irradiation data on the KWU/CE fuel rods in the project and discusses the failure behavior of PWR-fuel rods. (author)

  18. Fuel Rod Vibration Measurement Method using a Flap and its Verification

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Joo Young; Park, Nam Gyu; Suh, Jung Min; Jeon, Kyeong Lak [KEPCO NF Co., Daejeon (Korea, Republic of)

    2011-10-15

    Flow-induced vibration is a critical factor for the mechanical integrity of a fuel rod. This vibration can cause leaked fuel through the mechanism, such as grid to rod fretting. To minimize the failures caused by flow-induced vibration, a robust design is needed which takes into account vibrational characteristics. That is, the spacer grid design should be developed to avoid any excessive vibration. On the one hand, if fuel rod vibration can be measured, an estimation of the excitation forces, which are a critical cause of rod failure, should be possible. Therefore, by applying an external force, flow-induced vibration can be roughly estimated when the fuel rod vibration model is used. KEPCO Nuclear Fuel developed the test loop to research flow-induced vibration as shown in Fig.1. The investigation flow-induced vibration (INFINIT) - the test facility - can measure the grid strap vibration and pressure drop of a 5x5 small scale fuel bundle. Basically, using a Laser Doppler Vibrometer (LDV), the vibration of a structure immersed in high speed fluid can be measured. Grid strap vibration is easily measured using an LDV. However, it is quite difficult to measure fuel rod vibration because of the round surface shape of the rods. In addition, measuring current method using the LDV, it was only possible to directly measure fuel rod vibration at the first row of the bundle as the rods behind the first row are obscured. To solve this problem, a thin flap, as shown in Fig. 2(a) can be used as a reflecting target, gaining access to rods within the bundle. The flap is attached to the fuel rod, as in Fig. 2(b). As a result, most of the inner rod vibration can be measured. Before using a flap to measure fuel rod vibration, a verification process was needed to show whether the LDV signal from the flap vibration provided equivalent and reliable signals. Therefore, impact testing was carried out on the fuel rod using a flap. The LDV signals were then compared with accelerometer

  19. Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

    International Nuclear Information System (INIS)

    Vickerd, Meggan

    2013-01-01

    Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Site characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic

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

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    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)

  1. Theoretical investigations of the gas flow in ballooning LWR-fuel rods

    International Nuclear Information System (INIS)

    Gaballah, I.

    1978-09-01

    A theory is developed for the calculation of gas flow in a fuel rod simulator or in a fuel rod with round- or cracked pellets. The fundamental equations are formulated, simplified, reformed, and then numerically solved. The numerical investigations show, that a quasi steady incompressible flow model can be used without great error. The effect of the deformation form is studied. A uniform deformation along the whole length causes small pressure difference. A power profile and rod spacers cause non-uniform clad deformation of the fuel rod simulator or the fuel rod. This deformation leads to greater pressure differences. Finally the effect of the cracked pellets is studied. The cracked pellets cause great pressure differences along the fuel rod. (orig.) 891 HP [de

  2. Cost targets for at-reactor spent fuel rod consolidation

    International Nuclear Information System (INIS)

    Macnabb, W.V.

    1985-01-01

    The high-level nuclear waste management system in the US currently envisions the disposal of spent fuel rods that have been removed from their assemblies and reconfigured into closely packed arrays. The process of fuel rod removal and packaging, referred to as rod consolidation, can occur either at reactors or at an integrated packaging facility, monitored retrievable storage (MRS). Rod consolidation at reactors results in cost savings down stream of reactors by reducing needs for additional storage, reducing the number of shipments, and reducing (eliminating, in the extreme) the amount of fuel handling and consolidation at the MRS. These savings accrue to the nuclear waste fund. Although private industry is expected to pay for at-reactor activities, including rod consolidation, it is of interest to estimate cost savings to the waste system if all fuel were consolidated at reactors. If there are savings, the US Department of Energy (DOE) may find it advantageous to pay for at-reactor rod consolidation from the nuclear waste fund. This paper assesses and compares the costs of rod consolidation at reactors and at the MRS in order to determine at what levels the former could be cost competitive with the latter

  3. Nuclear fuel rod helium leak inspection apparatus and method

    International Nuclear Information System (INIS)

    Ahmed, H.J.

    1991-01-01

    This patent describes an inspection apparatus for testing nuclear fuel rods for helium leaks. It comprises a test chamber being openable and closable for receiving at least one nuclear fuel rod; means separate from the fuel rod for supplying helium and constantly leaking helium at a predetermined known positive value into the test chamber to constantly provide an atmosphere of helium at the predetermined known positive value in the test chamber; and means for sampling the atmosphere within the chamber and measuring the helium in the atmosphere such that a measured helium value below a preset minimum helium value substantially equal to the predetermined known positive value of the atmosphere of helium being constantly provided in the test chamber indicates a malfunction in the inspection apparatus, above a preset maximum helium value greater than the predetermined known positive in the test chamber indicates the existence of a helium leak from the fuel rod, or between the preset minimum and maximum helium values indicates the absence of a helium leak from the fuel rod

  4. Researches of WWER fuel rods behaviour under RIA accident conditions

    International Nuclear Information System (INIS)

    Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

    2003-01-01

    Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

  5. Measuring element for determining the internal pressure in fuel rods

    International Nuclear Information System (INIS)

    Deckers, H.; Drexler, H.; Reiser, H.

    1983-01-01

    A pressure cell is situated inside the fuel rod, which contains a magnetic core or a core influenced by magnetism, whose position relative to an outer front surface of an end stopper of the fuel rod can vary. The fuel rod contains a pressure cell directly above the lower end stopper or connected to it. This can consist of closed bellows, where if the internal pressure in the fuel rod rises, a ferrite core moves axially. When the pressure drops, this returns to the initial position, which is precisely defined by a stop. To detect a rod defect, the position of the soft iron core relative to the lower edge of the end stopper is scanned by a special measuring device. (orig./HP) [de

  6. Evolution of fuel rod support under irradiation consequences on the mechanical behavior of fuel assembly

    International Nuclear Information System (INIS)

    Billerey, A.; Bouffioux, P.

    2002-01-01

    The complete paper follows. According to the fuel management policy in French PWR with respect to high burn-up, the prediction of the mechanical behavior of the irradiated fuel assembly is required as far as excessive deformations of fuel assembly might lead to incomplete Rod Cluster Control Assembly insertion (safety problems) and fretting wear lead to leaking rods (plant operation problems). One of the most important parameter is the evolution of the fuel rod support in the grid cell as it directly governs the mechanical behavior of the fuel assembly and consequently allows to predict the behavior of irradiated structure in terms of (i) axial and lateral deformation (global behavior of the assembly) and (ii) fretting wear (local behavior of the rod). Fuel rod support is provided by a spring-dimple system fixed on the grid. During irradiation, the spring force decreases and a gap between the rod and the spring might open. This phenomenon is due to (i) irradiation-induced stress relaxation for the spring and for the dimples, (ii) grid growth and (iii) reduction of rod diameter. Two models have been developed to predict the behavior of the rod in the grid cell. The first model is able to evaluate the spring force relaxation during irradiation. The second one is able to evaluate the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (i) the creep laws of the grid materials, (ii) the growth law of the grid, (iii) the evolution of rod diameter and (iv) the design of the fuel rod support. The objectives of this paper are to: (i) evaluate the consequences of grid support design modifications on the fretting sensitivity in terms of predicted maximum gap during irradiation and operational time to gap appearance; (ii) evaluate, using a non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the mechanical behavior of the full assembly in terms of axial and

  7. Fuel-clad heat transfer coefficient of a defected fuel rod

    International Nuclear Information System (INIS)

    Bruet, M.; Stora, J.P.

    1976-01-01

    A special rod has been built with a stack of UO 2 pellets inside a thick zircaloy clad. The atmosphere inside the fuel rod can be changed and particularly the introduction of water is possible. The capsule was inserted in the Siloe pool reactor in a special device equipped with a neutron flux monitor. The fuel centerline temperature and the temperature at a certain radius of the clad were recorded by two thermocouples. The temperature profiles in the fuel and in the cladding have been calculated and then the heat transfer coefficient. In order to check the proper functioning of the device, two runs were successively achieved with a helium atmosphere. Then the helium atmosphere inside the fuel rod was removed and replaced by water. The heat transfer coefficients derived from the measurements at low power level are in agreement with the values given by the model based on thermal conductivity. However, for higher power levels, the heat transfer coefficients become higher than those based on the calculated gap

  8. Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

    1999-06-01

    Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

  9. System and method for consolidating spent fuel rods

    International Nuclear Information System (INIS)

    Baudro, T.O.

    1987-01-01

    A system is described for consolidating spent fuel rods from spent fuel assemblies, comprising: a consolidation container in which the fuel rods may be packed; a frame capable of holding a fuel assembly and the container during consolidation, the frame permitting each of the fuel assembly and the container to be removed; tool means with gripper means for gripping and releasing a rod, the tool means including means for moving the gripper means upwardly and downwardly; a first indexing head having first guide means for guiding the gripper means while the gripper means moves downwardly; a first rail, the first indexing head being slidably mounted on the first rail; a second indexing head having second guide means for guiding the gripper means while the gripper means moves downwardly; a second rail, the second indexing head being slidably mounted on the second rail; and a third rail, the first rail and the second rail being slidably mounted on the third rail; wherein the first indexing head is slidable on the first and third rails to a first position that is above a preselected rod in the fuel assembly; and wherein the second indexing head is slidable on the second and third rails to a second position that is above a preselected location in the container

  10. In-pile experiments on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Karb, E.; Pruessmann, M.; Sepold, L.

    1980-05-01

    This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

  11. Method of monitoring fuel-rod vibrations in a nuclear fuel reactor

    International Nuclear Information System (INIS)

    Kawamura, Makoto; Takai, Katsuaki.

    1985-01-01

    Purpose: To monitor the vibration modes of fuel rods continuously and on real time during operation of a PWR type nuclear reactor. Method: Vibrations of fuel rods during reactor operation are mainly caused by the lateral flow of coolants flowing through the gaps at the joints of reactor core buffle plates into a reactor core and fretting damages may possibly be caused to the fuel rod support portions due to the vibrations. In view of the above, self-powered detectors are disposed at a plurality of axial positions for the respective peripheral fuel assemblies in adjacent with the buffle plates and the detection signals from neutron detectors, that is, the fluctuations in neutrons are subjected to a frequency analysis during the operation period. The neutron detectors are disposed at the periphery of the reactor core, because the fuel assemblies disposed at the peripheral portion directly undergo the lateral flow from the joints of the buffle plates and vibrates most violently. Thus, the vibration situations can be monitored continuously, in a three demensional manner and on real time. (Moriyama, K.)

  12. Failed fuel rod detection system and computerized manipulator during outages

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1984-01-01

    During regular outages spent fuel assemblies need to be replaced and relocated within the core. Defective fuel rods in particular fuel assemblies have to be removed from further service and before delivery of such faulty fuel assemblies to a reprocessing plant. The system which Brown Boveri Reaktor GmbH and Krautkraemer have developed in the Federal Republic of Germany is capable of directly locating the defective rods in a proper fuel assembly. Inspection times are comparable to those of standard sipping methods, with the advantages of immediately available results and direct identification of the defective fuel rods. During the repair of fuel assemblies this system allows withdrawal of individual defective rods. With the sipping method all the fuel rods of a defective fuel assembly need to be removed and inspected by eddy current testing. During steam generator inspection and repair personnel are exposed to ample radiation. A remotely controlled, computerized manipulator was used to significantly reduce the radiation dose by automating steps in the procedures; at the same time inspection and repair times were reduced. The main features of the manipulator are a rigid component construction of the leg and two arms, and a resolver control for horizontal and vertical motion that enables rapid and accurate access to a desired tube (author)

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

    International Nuclear Information System (INIS)

    Lima, R. de C.F. de.

    1984-01-01

    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) [pt

  14. Fuel Rod Flow-Induced Vibration Overview

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kang, Heung Seok; Kim, Hyung Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    To ensure fuel design safety and structural integrity requires the response prediction of fuel rod to reactor coolant flow excitation. However, there are many obstacles in predicting the response as described. Even if the response can be predicted, the design criteria on wear failure, including correlation with the vibration, may be difficult to establish because of a variety of related parameters, such as material, surface condition and environmental factors. Thus, a prototype test for each new fuel assembly design, i.e. a long-term endurance test, is performed for design validation with respect to flow-induced vibration (FIV) and wear. There are still needs of theoretical prediction methods for the response and anticipated failure. This paper revisits the general aspect on the response prediction, mathematical description, analysis procedure and wear correlation aspect of fuel rod's FIV

  15. Evaluation of LWR fuel rod behavior under operational transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Hiramoto, K.; Maru, A.

    1984-01-01

    To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

  16. Remote helium leak test of the DUPIC fuel rod

    International Nuclear Information System (INIS)

    Kim, W. K; Kim, S. S.; Lim, S. P.; Lee, J. W.; Yang, M. S.

    1998-01-01

    DUPIC(Direct Use of spent PWR fuel In CANDU reactor) is one of dry reprocessing fuel cycles to reuse irradiated PWR fuel in CANDU power plant. DUPIC fuel is so radioactive that DUPIC fuel is remotely fabricated at hot cell such as IMEF hot cell in which radiation is shielded and remote operation is possible. In this study, Helium leakage has been tested for the simulated DUPIC fuel rod manufactured by Nd:YAG laser end-cap welding at simulated hot cell. The remote inspection technique has been developed to evaluate the soundness of DUPIC fuel fabricated through new processes. Vacuum chamber has been developed to be remotely operated by manipulators at hot cell. As the result of remote test, Helium leakage of DUPIC fuel rod is around background level, CANDU specification has been satisfied. In the result of the study, remote test has been successfully performed at the simulated hot cell, and the soundness of DUPIC fuel rod welded by Nd:YAG laser has been confirmed

  17. Nuclear reactor internals construction and failed fuel rod detection system

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

  18. Critical experiments supporting underwater storage of tightly packed configurations of spent fuel rods

    International Nuclear Information System (INIS)

    Hoovler, G.S.; Baldwin, M.N.

    1981-04-01

    Criticla arrays of 2.5%-enriched UO 2 fuel rods that simulate underwater rod storage of spent power reactor fuel are being constructed. Rod storage is a term used to describe a spent fuel storage concept in which the fuel bundles are disassembled and the rods are packed into specially designed cannisters. Rod storage would substantially increase the amount of fuel that could be stored in available space. These experiments are providing criticality data against which to benchmark nuclear codes used to design tightly packed rod storage racks

  19. IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    1996-01-01

    Description: It contains data from experiments that have been performed at the IFE/OECD Halden Reactor Project, available for use in fuel performance studies. It covers experiments on thermal performance, fission product release, clad properties and pellet clad mechanical interaction. It includes also experimental data relevant to high burn-up behaviour. IFA-432: Measurements of fuel temperature response, fission gas release and mechanical interaction on BWR-type fuel rods up to high burn-ups. The assembly featured several variations in rod design parameters, including fuel type, fuel/cladding gap size, fill gas composition (He and Xe) and fuel stability. It contained 6 BWR-type fuel rods with fuel centre thermocouples at two horizontal planes, rods were also equipped with pressure transducers and cladding extensometers. Only data from 6 rods are compiled here

  20. Reactor core with rod-shaped fuel cells

    International Nuclear Information System (INIS)

    Dworak, A.

    1975-01-01

    Power distribution in a high-temperature gas-cooled reactor is optimized. Especially the axial as well as the radial power distribution is kept constant, the core consisting of several consecutive rod-shaped fuel cells. To this end, the dwell times of the fuel cells are fitted to the given power distribution. Fuel cells with equal dwell times, seen in flow direction, are arranged side by side, and those with the shortest dwell times are placed in areas with the greatest power release. These areas ly on the coolant inlet side. To keep the power distribution constant, fuel cells with neutron poison or absorber rods with absorbing rates decreasing in flow direction can also be inserted. (RW/PB) [de

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

    International Nuclear Information System (INIS)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

    2013-01-01

    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

  2. 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.

  3. Storage device for fuel rods of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Kempf, B.

    1983-01-01

    The storage device, which can be flexibly matched to the number of fuel rods to be stored and is not tied to a space, has a vertical support post situated on the floor and a stiff upright also situated vertically on the floor, which is used to accommodate at least one fuel rod. The stiff upright is connected directly to the support post by connections which can be undone, or form locking via another vertical stiff upright situation on the floor. (orig./HP) [de

  4. Methodology of fuel rod design for pressurized light water reactors

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  5. Model investigation of fuel rod behaviour

    International Nuclear Information System (INIS)

    Girgis, M.M.; Wiesenack, W.; Stegemann, D.

    1985-06-01

    Thermal and mechanical behaviour of fuel rods can be explained but unsatisfactorily by models based of an axial symmetry concept. Recently developed models include, with respect to their thermal components, a simple method for the computation of the temperature distribution within the fuel, and they also take into account the influence of excentrically placed pellets for the computation of heat transfer in the cold gap. Additionally, a finite-element model is used to evaluate the effects of cracking and fragmentation on the thermal behaviour of pellets. The reaction of fuel and fuel cladding to external and internal loadings and the axial interaction between fuel and cladding are described in the mechanical portion of the model. A special case of axial coupling is the so-called random stacking interaction caused by fuel pellets placed excentrically at the cladding and sliding radially and axially. In the comparison of measurement results, both thermal and mechanical behaviour of different rods from the OECD Halden Reactor Project are subject to investigations. (RF) [de

  6. On Cherenkov light production by irradiated nuclear fuel rods

    International Nuclear Information System (INIS)

    Branger, E.; Grape, S.; Svärd, S. Jacobsson; Jansson, P.; Sundén, E. Andersson

    2017-01-01

    Safeguards verification of irradiated nuclear fuel assemblies in wet storage is frequently done by measuring the Cherenkov light in the surrounding water produced due to radioactive decays of fission products in the fuel. This paper accounts for the physical processes behind the Cherenkov light production caused by a single fuel rod in wet storage, and simulations are presented that investigate to what extent various properties of the rod affect the Cherenkov light production. The results show that the fuel properties have a noticeable effect on the Cherenkov light production, and thus that the prediction models for Cherenkov light production which are used in the safeguards verifications could potentially be improved by considering these properties. It is concluded that the dominating source of the Cherenkov light is gamma-ray interactions with electrons in the surrounding water. Electrons created from beta decay may also exit the fuel and produce Cherenkov light, and e.g. Y-90 was identified as a possible contributor to significant levels of the measurable Cherenkov light in long-cooled fuel. The results also show that the cylindrical, elongated fuel rod geometry results in a non-isotropic Cherenkov light production, and the light component parallel to the rod's axis exhibits a dependence on gamma-ray energy that differs from the total intensity, which is of importance since the typical safeguards measurement situation observes the vertical light component. It is also concluded that the radial distributions of the radiation sources in a fuel rod will affect the Cherenkov light production.

  7. DANCOFF-3, Dancoff Correction for Cylindrical Fuel Rod at H2O Gaps and for Fuel Clusters

    International Nuclear Information System (INIS)

    Sauer, A.

    1989-01-01

    1 - Nature of physical problem solved: Calculation of the Dancoff correction for cylindrical fuel rods in square and hexagonal infinite lattices, for fuel element rods near water gaps, and for fuel rod clusters. 2 - Method of solution: Evaluation by direct numerical integration over the moderator region. 3 - Restrictions on the complexity of the problem: For every rod arrangement at most 100 cases with different materials cross- sections

  8. Nuclear reactor fuel rod behavior modelling and current trends

    International Nuclear Information System (INIS)

    Colak, Ue.

    2001-01-01

    Safety assessment of nuclear reactors is carried out by simulating the events to taking place in nuclear reactors by realistic computer codes. Such codes are developed in a way that each event is represented by differential equations derived based on physical laws. Nuclear fuel is an important barrier against radioactive fission gas release. The release of radioactivity to environment is the main concern and this can be avoided by preserving the integrity of fuel rod. Therefore, safety analyses should cover an assessment of fuel rod behavior with certain extent. In this study, common approaches for fuel behavior modeling are discussed. Methods utilized by widely accepted computer codes are reviewed. Shortcomings of these methods are explained. Current research topics to improve code reliability and problems encountered in fuel rod behavior modeling are presented

  9. COMETHE III-M for transient fuel rod behaviour prediction

    International Nuclear Information System (INIS)

    Billaux, M.; Vliet, J. van

    1983-01-01

    The COMETHE III-M version is being developed in order to provide fuel rod behaviour prediction capability both in steady-state and in transient situations. It also allows to estimate the fuel rod enthalpy evolution versus time or burnup which may be important in core-related safety studies. This paper describes the transient heat transfer models, including transient heat conduction inside the fuel rod, and a subchannel model providing transient flow as well as enthalpy calculation capability. Transient fission gas release is also modelled on basis of the change rate of oxide temperature. The models are illustrated by a few calculation examples. (author)

  10. The buckling of fuel rods in transportation casks under hypothetical accident conditions

    International Nuclear Information System (INIS)

    Bjorkman, G.S.

    2004-01-01

    The buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations following a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding the higher the inertia loads on the cladding, and, therefore, the lower the ''g'' value at which buckling occurs. Current published solutions do not consider displacement compatibility between the fuel and the cladding. By invoking displacement compatibility between the fuel column and the cladding column, this paper presents an exact solution for the buckling of fuel rods under inertia loading. The results show that the critical inertia load magnitude for the buckling of a fuel rod depends on the weight of the cladding and the total weight of the fuel, regardless of the percentage of fuel mass that is assumed to be attached to or participate with the cladding in the buckling process. Therefore, 100% of the fuel always participates in the buckling of a fuel rod under inertia loading

  11. The nuclear fuel rod character recognition system based on neural network technique

    International Nuclear Information System (INIS)

    Kim, Woong-Ki; Park, Soon-Yong; Lee, Yong-Bum; Kim, Seung-Ho; Lee, Jong-Min; Chien, Sung-Il.

    1994-01-01

    The nuclear fuel rods should be discriminated and managed systematically by numeric characters which are printed at the end part of each rod in the process of producing fuel assembly. The characters are used to examine manufacturing process of the fuel rods in the inspection process of irradiated fuel rod. Therefore automatic character recognition is one of the most important technologies to establish automatic manufacturing process of fuel assembly. In the developed character recognition system, mesh feature set extracted from each character written in the fuel rod is employed to train a neural network based on back-propagation algorithm as a classifier for character recognition system. Performance evaluation has been achieved on a test set which is not included in a training character set. (author)

  12. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 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

  13. Fuel rod for use in BWR type reactor

    International Nuclear Information System (INIS)

    Takeuchi, Kiyoshi.

    1989-01-01

    A hollow intermediate end plug is disposed to a plenum portion of a fuel rod and a plenum spring is disposed between the end plug and the upper end of a fuel pellet. Then, a hollow portion is disposed between the intermediate end plug and an upper end plug. Thus, since a only a non exothermic portion is present from the intermediate end plug to the upper end plug, oxidation, corrosion, etc. to the fuel can are not caused so much as in the exothermic portion. Accordingly, the wall thickness of the fuel may be reduced to such a extent as only capable of withstanding the external pressure by coolants and the increasing inner pressure due to the release of FP gases and, accordingly, the wall thickness can be reduced as compared with that of the fuel portion in the fuel can. Further, since the power density per unit length of the fuel rod is reduced for fuels with increased number of fuel rods, it is possible to design so as to reduce the release amount of FP gases thereby decreasing the plenum volume. Further, since the surface area in the coolant phase stream portion is reduced, it can be expected for decreasing the pressure loss of fuels and accompanying effect for improving the channel stability. (T.M.)

  14. 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.

  15. Radiography inspection of weld for nuclear fuel rod

    International Nuclear Information System (INIS)

    Zhang Kai; Zhang Xichang

    1995-05-01

    The survey of radiography inspection, advantages, disadvantages and applications of main kinds of radiography inspection methods are presented. Emphasis is put upon the structure and functions of X-ray flaw detecting device for nuclear fuel rod welds, the actuating program of the device, as well as the structure of some key mechanism and the functions of them. The analysis is made upon the actuating principles. Finally, the test of long-term operation proves the device to be stable in operation, reliable in action, to possess high level of automation and high sensitivity and it can simultaneously perform on-line X-ray inspection of 25 nuclear fuel rods with a diameter less than 10 mm, and meet the requirements of large-scale production of nuclear fuel rods (5 figs.)

  16. Mechanical behaviour of PWR fuel rods during intermediate storage

    International Nuclear Information System (INIS)

    Bouffioux, P.; Dalmas, R.; Bernaudat, C.

    2000-01-01

    EDF, which owns the irradiated fuel coming from its NPPs, has initiated studies regarding the mechanical behaviour of a fuel rod and the integrity of its cladding, in the case where the spent fuel is stored for a significant duration. During the phases following in-reactor irradiation (ageing in a water-pool, transport and intermediate storage), many phenomena, which are strongly coupled, may influence the cladding integrity: - residual power and temperature decay; - helium production and release in the free volume of the rod (especially for MOX fuel); - fuel column swelling; - cladding creep-out under the inner gas pressure of the fuel rod; - metallurgical changes due to high temperatures during transportation. In parallel, the quantification of the radiological risk is based on the definition of a cladding integrity criterion. Up to now, this criterion requires that the clad hoop strain due to creep-out does not exceed 1%. A more accurate criterion is being investigated. The study and modelling of all the phenomena mentioned above are included in a R and D programme. This programme also aims at redefining the cladding integrity criterion, which is assumed to be too conservative. The R and D programme will be presented. In order to predict the overall behaviour of the rod during the intermediate storage phases, the AVACYC code has been developed. It includes the models developed in the R and D programme. The input data of the AVACYC code are provided by the results of in-reactor rod behaviour simulations, using the thermal-mechanical CYRANO3 code. Its main results are the evolution vs. time of hoop stresses in the cladding, rod internal pressure and cladding hoop strains. Chained CYRANO-AVACYC calculations have been used to simulate the behaviour of MOX fuel rods irradiated up to 40 GWd/t and stored under air during 100 years, or under water during 50 years. For such fuels, where the residual power remains high, we show that a large part of the cladding strain

  17. Express diagnostics of WWER fuel rods at nuclear power plants

    International Nuclear Information System (INIS)

    Pavlov, S.; Amosov, S.; Sagalov, S.; Kostyuchenko, A.

    2009-01-01

    Higher safety and economical efficiency of nuclear power plants (NPP) call for a continuous design modification and technological development of fuel assemblies and fuel rods as well as optimization of their operating conditions. In doing so the efficiency of new fuel introduction depends on the completeness of irradiated fuel data in many respects as well as on the rapidity and cost of such data obtaining. Standard examination techniques of fuel assemblies (FA) and fuel rods (FR) intended for their use in hot cell conditions do not satisfy these requirements in full extent because fuel assemblies require preliminary cooling at NPP to provide their shipment to the research center. Expenditures for FA transportation, capacity of hot cells and expenditures for the examined fuel handling do not make it possible to obtain important information about the condition of fuel assemblies and fuel rods after their operation. In order to increase the comprehensiveness of primary data on fuel assemblies and fuel rods immediately after their removal from the reactor, inspection test facilities are widely used for these purposes. The inspection test facilities make it possible to perform nondestructive inspection of fuel in the NPP cooling pools. Moreover these test facilities can be used to repair failed fuel assemblies. The ultrasonic testing of failed fuel rods inside the fuel assembly was developed for stands of inspection and repair of TVSA WWER-1000 for the Kalinin NPP and Temelin NPP. This method was tested for eight leaking fuel assemblies WWER-440 and WWER-1000 with a burnup of ∼14 up to 38 MW·day/kgU. The ultrasonic testing proved its high degree of reliability and efficiency. The defectoscopy by means of the pulsed eddy-current method was adapted for the stand of inspection and repair of TVSA WWER-1000 for the Kalinin NPP. This method has been used at RIAR as an express testing method of FR claddings during the post-irradiation examinations of fuel assemblies WWER

  18. Fuel rod failure detection method and system

    International Nuclear Information System (INIS)

    Assmann, H.; Janson, W.; Stehle, H.; Wahode, P.

    1975-01-01

    The inventor claims a method for the detection of a defective fuel rod cladding tube or of inleaked water in the cladding tube of a fuel rod in the fuel assembly of a pressurized-water reactor. The fuel assembly is not disassembled but examined as a whole. In the examination, the cladding tube is heated near one of its two end plugs, e.g. with an attached high-frequency inductor. The water contained in the cladding tube evaporates, and steam bubbles or a condensate are detected by the ultrasonic impulse-echo method. It is also possible to measure the delay of the temperature rise at the end plug or to determine the cooling energy required to keep the end plug temperature stable and thus to detect water ingression. (DG/AK) [de

  19. Fuel element cellular grid structure and procedure to insert and withdraw fuel rods from that structure

    International Nuclear Information System (INIS)

    1975-01-01

    A typical embodiment of the invention provides a means for selectively inserting and withdrawing one or more fuel rods from a fuel element cellular grid structure. The transverse stubs on one side of a long, thin bar are turned through 90deg to extend across the gap between mutually perpendicular grid structure plates. The extreme ends of these stubs engage the adhacent portions of the associated plates that form part of the grid cells. Pressing the stubs against the plate portions through the application of appropriate force in a longitudinal direction relative to the bar deflects the engaged plates through a sufficient distance to enable fuel rods to be inserted into or withdrawn from respective cells. After rod insertion, the force applied to the bar is released to enable the plates to relax and engage the fuel rods. The bars are rotated once more through 90deg and withdrawn from the grid structure. A similar procedure is employed to withdraw fuel rods from the grid structure

  20. Status and development of RBMK fuel rods and reactor materials

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Reshetnikov, F.G.; Ioltukhovsky, A.G.

    1998-01-01

    The paper presents current status and development of RBMK fuel rods and reactor materials. With regard to fuel rod cladding the following issues have been discussed: corrosion, tensile properties, welding technology and testing of an alternative cladding alloy with a composition of Zr-Nb-Sn-Fe. Erbium doped fuel has been suggested for safety improvement. Also analysis of fuel reliability is presented in the paper. (author)

  1. Determination of the perturbing effect of the measuring device on thermal neutron distribution inside the fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Takac, S M; Krcevinac, S B [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-07-15

    The measurement of the thermal neutron distribution in an elementary cell of the reactor core is based on activating some of the existing detectors such as gold, copper, dysprosium, etc., inside the fuel rod and the corresponding part of the slowing-down medium. The techniques of measuring may be classified in two groups: - technique with detector foils, and - technique with a detector wire. The first group includes all experiments based on the so called 'tube technique'. By this technique the detector foils are arranged specifically in the tube by means of spacers and are positioned in a radially bored fuel rod. The 'spiral technique' is also included here. By this technique the fuel rod, which is first radially cut, is axially bored along the spiral and then detector foils inserted in the holes. The second group includes techniques according to which the detector wires may be positioned either in the radially bored hole through the fuel rod or in the spiral groove made in the horizontal cross-section in the fuel rod. To obtain higher resolution the detector wire after activation can be extruded, or before irradiation, spirally wound around a solid core and thus positioned in the radial hole in the fuel rod. In all cases the fuel region is perturbed either by the holes and the detector material, or by the holder of the detector foils. A number of authors have carried out these experiments under different geometrical and nuclear conditions, so obviously this perturbation had different effects on the results. So far the cell perturbation effects have not been discussed in the literature, neither this effect has been corrected in the final results. With respect to this a series of experiments for determining the micro distribution of thermal neutrons inside the fuel rod were made on the heavy-water natural-uranium system for different lattice pitches, with special stress or the investigation of the perturbing effects in the fuel rod which inevitably must be

  2. Determination of the perturbing effect of the measuring device on thermal neutron distribution inside the fuel rod

    International Nuclear Information System (INIS)

    Takac, S.M.; Krcevinac, S.B.

    1966-07-01

    The measurement of the thermal neutron distribution in an elementary cell of the reactor core is based on activating some of the existing detectors such as gold, copper, dysprosium, etc., inside the fuel rod and the corresponding part of the slowing-down medium. The techniques of measuring may be classified in two groups: - technique with detector foils, and - technique with a detector wire. The first group includes all experiments based on the so called 'tube technique'. By this technique the detector foils are arranged specifically in the tube by means of spacers and are positioned in a radially bored fuel rod. The 'spiral technique' is also included here. By this technique the fuel rod, which is first radially cut, is axially bored along the spiral and then detector foils inserted in the holes. The second group includes techniques according to which the detector wires may be positioned either in the radially bored hole through the fuel rod or in the spiral groove made in the horizontal cross-section in the fuel rod. To obtain higher resolution the detector wire after activation can be extruded, or before irradiation, spirally wound around a solid core and thus positioned in the radial hole in the fuel rod. In all cases the fuel region is perturbed either by the holes and the detector material, or by the holder of the detector foils. A number of authors have carried out these experiments under different geometrical and nuclear conditions, so obviously this perturbation had different effects on the results. So far the cell perturbation effects have not been discussed in the literature, neither this effect has been corrected in the final results. With respect to this a series of experiments for determining the micro distribution of thermal neutrons inside the fuel rod were made on the heavy-water natural-uranium system for different lattice pitches, with special stress or the investigation of the perturbing effects in the fuel rod which inevitably must be

  3. Transition of Natural Frequencies of a Fuel Rod during Its Lifetime

    International Nuclear Information System (INIS)

    Kim, Hyeong Koo; Lee, Kyou Seok; Kim, Jeong Ha; Jeon, Sang Yoon

    2009-01-01

    The natural frequencies of a Pressurized Water Reactor (PWR) fuel rod are dependent on the geometrical and mechanical properties of fuel rod itself and its supporting conditions provided by spacer grids. By the way, these environmental parameters suffer remarkable change due to the plant operating conditions such as burnup, temperature, system pressure, and so on. It is inevitable, therefore, to be changed the natural frequencies of the fuel rod during its lifetime. In this paper, the transition of natural frequencies of the fuel rod for OPR1000 plants has been investigated considering fuel conditions associated with fuel life time. Basically for this investigation, three analysis models have been proposed representing beginning-of life (BOL) condition, middle-of-life (MOL) condition and end-of-life (EOL) condition including spacer grid supporting conditions. With these models, several modal analyses have been performed and the results have been compared with those of the test which has been carried out for verification of the analysis model. With these analyses and test, the changing vibration behavior of the PLUS7 fuel rod for OPR1000 during its life time has been discussed

  4. Optimization of in-core fuel management and control rod strategy in equilibrium fuel cycle

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1975-01-01

    An in-core fuel management problem is formulated for the equilibrium fuel cycle in an N-region nuclear reactor model. The formulation shows that the infinite multiplication factor k infinity requisite for newly charged fuel can be separated into two terms - one corresponding to the average k infinity at the end of the cycle and the other representing the direct contribution of the shuffling scheme and control rod programming. This formulation is applied to a three-region cylindrical reactor to obtain simultaneous optimization of shuffling and control rod programming. It is demonstrated that this formulation aids greatly in gaining a better understanding of the effects of changes in the shuffling scheme and control rod programming on equilibrium fuel cycle performance. (auth.)

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

    International Nuclear Information System (INIS)

    Rushton, J.E.; Allen, E.J.; Chiles, M.M.; Jenkins, J.D.

    1979-11-01

    Refabricated HTGR fuel rods will contain from approx. 0.15 to 0.5 g 233 U and/or 235 U. 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 232 U 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. Mechanical stress analysis for a fuel rod under normal operating conditions

    International Nuclear Information System (INIS)

    Pino, Eddy S.; Giovedi, Claudia; Serra, Andre da Silva; Abe, Alfredo Y.

    2013-01-01

    Nuclear reactor fuel elements consist mainly in a system of a nuclear fuel encapsulated by a cladding material subject to high fluxes of energetic neutrons, high operating temperatures, pressure systems, thermal gradients, heat fluxes and with chemical compatibility with the reactor coolant. The design of a nuclear reactor requires, among a set of activities, the evaluation of the structural integrity of the fuel rod submitted to different loads acting on the fuel rod and the specific properties (dimensions and mechanical and thermal properties) of the cladding material and coolant, including thermal and pressure gradients produced inside the rod due to the fuel burnup. In this work were evaluated the structural mechanical stresses of a fuel rod using stainless steel as cladding material and UO 2 with a low degree of enrichment as fuel pellet on a PWR (pressurized water reactor) under normal operating conditions. In this sense, tangential, radial and axial stress on internal and external cladding surfaces considering the orientations of 0 deg, 90 deg and 180 deg were considered. The obtained values were compared with the limit values for stress to the studied material. From the obtained results, it was possible to conclude that, under the expected normal reactor operation conditions, the integrity of the fuel rod can be maintained. (author)

  7. Pulsed eddy current inspection system for nondestructive examination of irradiated fuel rods

    International Nuclear Information System (INIS)

    Yancey, M.E.

    1979-01-01

    An inspection system has been developed for nondestructive examination of irradiated fuel rods utilizing pulsed eddy current techniques. The system employs an encircling type pulsed eddy current transducer capable of sensing small defects located on both the inner and outer diameter fuel rod surfaces during a single scan. Pulsed eddy current point probes are used to provide fuel rod wall thikness data and an indication of radial defect location. Two linear variable differential transformers are used to provide information on fuel rod diameter variation. A microprocessor based control system is used to automatically scan fuel rods up to 4.06 meters in length at predetermined radial locations. Defects as small as 0.005 cm deep by 0.254 cm long by 0.005 cm wide have been detected on outside diameter surfaces of a 1.43 cm outside diameter fuel rod cladding with a 0.094 cm wall thickness and 0.010 cm deep by 0.254 cm long by 0.005 cm wide on the inside diameter surface

  8. Pressure equalization systems in pressurized water reactor fuel rods

    International Nuclear Information System (INIS)

    Steven, J.; Wunderlich, F.

    1979-01-01

    For the development of a pressure reduction system in PWR fuel rods the capability of charcoal to adsorb Helium, Xenon and Krypton at temperatures of about 300 0 C was investigated. The influence of the adsorption on fuel rod internal pressure and in creep strain on the tube was evaluated in a design study. (orig.) [de

  9. Study of pellet clad interaction defects in Dresden-3 fuel rods

    International Nuclear Information System (INIS)

    Pasupathi, V.; Perrin, J.S.

    1979-01-01

    During Cycle-3 operation of Dresden-3, fuel rod failures occurred following a transient power increase. Ten fuel rods from five of the leaking fuel assemblies were examined at Battelle's Columbus Laboratory and General Electric-Vallecitos Nuclear Center. Examinations consisted of nondestructive and destructive methods including metallography and scanning electron microscopy (SEM). Results showed the cause of fuel rod failure to be pellet clad interaction involving stress corrosion cracking. Results of SEM studies of the cladding crack surfaces and deposits on clad inner surfaces were in agreement with those reported by other investigators

  10. Experiments with preirradiated fuel rods in the Nuclear Safety Research Reactor

    International Nuclear Information System (INIS)

    Horiki, O.; Kobayashi, S.; Takariko, I.; Ishijima, K.

    1992-01-01

    In the Nuclear Safety Research Reactor (NSRR) owned and operated by Japan Atomic Energy Research Institute (JAERI), extensive experimental studies on the fuel behavior under reactivity initiated accident (RIA) conditions have been continued since the start of the test program in 1975. Accumulated experimental data were used as the fundamental data base of the Japanese safety evaluation guideline for reactivity initiated events in light water cooled nuclear power plants established by the nuclear safety commission in 1984. All of the data used to establish the guideline were, however, limited to those derived from the tests with fresh fuel rods as test samples because of the lack of experimental facility to handle highly radioactive materials.The guideline, therefore, introduces the peak fuel enthalpy of 85 cal/g which was adopted from the SPERT-CDC data as a provisional failure threshold of preirradiated fuel rod and, says that this value should be revised based on the NSRR experiments in the future. According to the above requirement, new NSRR experimental program with the preirradiated fuel rods as test samples was started in 1989. Test fuel rods are prepared by refabrication of the long-sized fuel rods preirradiated in commercial PWRs and BWRs into short segments and by preirradiation of short-sized test fuel rods in the Japan Material Testing Reactor(JMTR). For the tests with preirradiated fuel rods as test samples, the special experimental capsules, the automatic instrumentation fitting device, the automatic capsule assembling device and the capsule loading device were newly developed. In addition, the existing hot cave was modified to mount the capsule assembling device and the other inspection tools and, a new small iron cell was established adjacent to the cave to store the instrumentation fitting device. (author)

  11. Status of work on the final repository concept concerning direct disposal of spent fuel rods in fuel rod casks (BSK)

    International Nuclear Information System (INIS)

    Filbert, W.; Wehrmann, J.; Bollingerfehr, W.; Graf, R.; Fopp, S.

    2008-01-01

    The reference concept in Germany on direct final storage of spent fuel rods is the burial of POLLUX containers in the final repository salt dome. The POLLUX container is self-shielded. The final storage concept also includes un-shielded borehole storage of high-level waste and packages of compacted waste. GNS has developed a spent fuel container (BSK-3) for unshielded borehole storage with a mass of 5.2 tons that can carry the fuel rods of three PWR reactors of 9 BWR reactors. The advantages of BSK storage include space saving, faster storage processes, less requirements concerning technical barriers, cost savings for self-shielded casks.

  12. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    International Nuclear Information System (INIS)

    Vitkova, M.; Manolova, M.; Stefanova, S.; Simeonova, V.; Passage, G.; Lassmann, K.

    1994-01-01

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: 1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; 2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; 3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs

  13. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    Energy Technology Data Exchange (ETDEWEB)

    Vitkova, M; Manolova, M; Stefanova, S; Simeonova, V; Passage, G [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Kharalampieva, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Lassmann, K [European Atomic Energy Community, Karlsruhe (Germany). European Inst. for Transuranium Elements

    1994-12-31

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: (1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; (2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; (3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs.

  14. Modelling of pellet cladding interaction during power ramps in PWR rods by means of Transuranus fuel rod analysis code

    International Nuclear Information System (INIS)

    Di Marcello, V.; Luzzi, L.

    2008-01-01

    Pellet-cladding interaction (PCI) in PWR type rods subjected to power ramps was analysed by means of TRANSURANUS (TU) fuel rod performance code. PCI phenomena depend on the fuel power history - i.e. by several irradiation and thermal induced phenomena occurring in the fuel rod and mutually interacting during its life in reactor - and may become critical for cladding integrity under accidental conditions. Ten test fuel rods, whose power histories and post irradiation experiment (PIE) data were available from the OECD/NEA-IAEA International Fuel Performance Experiment (UTE) database through the Studsvik SUPER-RAMP Project, were simulated by TRANSURANUS. During a power ramp pellet gaseous swelling can be inhibited by cladding pressure and can be over-predicted by a normal operation swelling model. This phenomenon was simulated by a new formulation of a fuel swelling model already available in the code, in order to consider hot pressing of inter-granular -fuel porosity due to the high hydrostatic stress resulting from PCI: it was found that TRANSURANUS, as a result of the proposed swelling formulation as well as of the accurate modelling of the other phenomena occurring during irradiation, gives correct predictions on PCI induced fuel rod failures. In addition, PCI failure threshold identified by TRANSURANUS was compared with the technological limits known in literature: the possibility of relaxing these limits for low burn-up values and the preponderance of the European fuel rod design in front of PCI emerged from TU analyses. Finally, a good agreement was found between TU evaluations and PIE data, with regard to fission gas release, fuel grain growth, and creep, corrosion and elongation of the cladding. (authors)

  15. Fuel assembly and burnable poison rod

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1993-01-01

    In a fuel assembly having burnable poison rods arranged therein, the burnable poison comprises an elongate small outer tube and an inner tube coaxially disposed within the outer tube. Upper and lower end tubes each sealed at one end are connected to both of the upper and lower ends in the inner and the outer tubes respectively. A coolant inlet hole is disposed to the lower end tube, while a coolant leakage hole is disposed to the upper end tube. Burnable poison members are filled in an annular space. Further, the burnable poison-filling region is disposed excepting portions for 1/20 - 1/12 of the effective fuel length at each of the upper and the lower ends of the fuel rod. Then, the concentration of the burnable poisons in a region above a boundary defined at a position 1/3 - 1/2, from beneath, of the effective fuel length is made smaller than that in the lower region. This enables to suppress excess reactions of fuels to reduce the mass of the burnable neutron. Excellent reactivity control performance at the initial stage of the burning can be attained. (T.M.)

  16. Inspection device for fuel rod restraint by support lattice of fuel assembly

    International Nuclear Information System (INIS)

    Hasegawa, Isao; Senga, Masatoshi; Kada, Mitoshi.

    1991-01-01

    An inspection operation section for disposing fuel assembly vertically at predetermined positions, a control section wired therewith, a moving operation section movable in the direction of X, Y and Z axes by a driving signal sent from the control section are disposed to an inspection section main body. A downward bore scope and a upward bore scope, each of such a size as can be inserted to the gaps between the fuel rods, are disposed while opposing to each other for observing the inside of each of cells from above and below in support lattices of fuel assemblies. High performance television cameras are disposed to each of bore scopes to supply images to monitoring televisions in the control section. Thus, a displacing operation section of the inspection operation section is automatically controlled three-dimensionally, the downward bore scope and the upward bore scope are integrally intruded to the inside of the gaps between the predetermined fuel rods from a required height and stopped at a predetermined position, mounted automatically to a required cell of the support lattice to efficiently observe and inspect the fuel rod restraint. (N.H.)

  17. Effects of fuel relocation on reflood in a partially-blocked rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Jae [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 34134 (Korea, Republic of); Kim, Jongrok; Kim, Kihwan; Bae, Sung Won [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Ballooning of the fuel rods has been an important issue, since it can influence the coolability of the rod bundle in a large-break loss-of-coolant accident (LBLOCA). Numerous past studies have investigated the effect of blockage geometry on the heat transfer in a partially blocked rod bundle. However, they did not consider the occurrence of fuel relocation and the corresponding effect on two-phase heat transfer. Some fragmented fuel particles located above the ballooned region may drop into the enlarged volume of the balloon. Accordingly, the fuel relocation brings in a local power increase in the ballooned region. The present study’s objective is to investigate the effect of the fuel relocation on the reflood under a LBLOCA condition. Toward this end, experiments were performed in a 5 × 5 partially-blocked rod bundle. Two power profiles were tested: one is a typical cosine shape and the other is the modified shape considering the effect of the fuel relocation. For a typical power shape, the peak temperature in the ballooned rods was lower than that in the intact rods. On the other hand, for the modified power shape, the peak temperature in the ballooned rods was higher than that in the intact rods. Numerical simulations were also performed using the MARS code. The tendencies of the peak clad temperatures were well predicted.

  18. The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, Y. K.; Kim, E. K.

    1996-01-01

    The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

  19. Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

    Science.gov (United States)

    Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

    2018-02-01

    SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

  20. SEFLEX - fuel rod simulator effects in flooding experiments. Pt. 2

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from unblocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5 x 5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5 x 5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  1. SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from blocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5x5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5x5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  2. Neural signal processing for identifying failed fuel rods in nuclear reactors

    International Nuclear Information System (INIS)

    Seixas, Jose M. de; Soares Filho, William; Pereira, Wagner C.A.; Teles, Claudio C.B.

    2002-01-01

    Ultrasonic pulses were used for automatic detection of failed nuclear fuel rods. For experimental tests of the proposed method, an assembly prototype of 16 x 16 rods was built by using genuine rods but without fuel inside (just air). Some rods were partially filled with water to simulate cracked rods. Using neural signal processing on the received echoes of the emitted ultrasonic pulses, a detection efficiency of 97% was obtained. Neural detection is shown to outperform other classical discriminating methods and can also reveal important features of the signal structure of the received echoes. (author)

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

    International Nuclear Information System (INIS)

    Estevez, Esteban; Markiewicz, Mario; Marino, Armando

    2000-01-01

    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. Process and equipment for pressure build-up in nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Heer, W.F.; Carli, E.V. de.

    1976-01-01

    The equipment makes possible the build-up of inert gas pressure in a filled and closed fuel can, i.e. in a complete fuel rod. Handling is simple, it is suitable for mass production and only causes low processing costs. The quality, e.g. the degree of purity of the contents of the rod, remains unchangedin processing. The equipment consists of a vacuum-tight space, into which the equally vacuum tight fuel rod is introduced, and can be fixed so that its position can be reproduced unmistakeably. The vacuum space contains a connection for the inert gases and a laser arrangement. After inserting a fuel rod into the facility, this is evacuated and the fuel can has a hole bored in it by a laser beam. After fast equalisation of pressure, an inert gas at the required pressure is introduced into the chamber and the fuel rod. After the filling process is completed, the fuel can is closed again with the same laser beam. The quality of the seal obtained, i.e the leak-tightness of the fuel can, can be checked after reduction of the inert gas pressure and before taking out the fuel rod, by repeated evacuation of the chamber. Laser light energies between 13,000 and 110,000 Joule/sq cm are sufficient. Optimum results were obtained for a Zircaloy fuel can with about 52,000 Joule/sq cm. (TK) [de

  5. Integration of post-irradiation examination results of failed WWER fuel rods

    International Nuclear Information System (INIS)

    Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Perepelkin, S.

    2003-01-01

    The aim of the work is to investigate the causes of WWER fuel rod failures and to reveal the dependence of the failed fuel rod behaviour and state on the damage characteristics and duration of their operation in the core. The post-irradiation examination of 12 leaky fuel assemblies (5 for WWER-440 and 7 for WWER-1000) has been done at SSC RF RIAR. The results show that the main mechanism responsible for the majority of cases of the WWER fuel rod perforation is debris-damage of the claddings. Debris fretting of the claddings spread randomly over the fuel assembly cross-section and they are registered in the area of the bundle supporting grid or under the lower spacer grids along the fuel assembly height. In the WWER fuel rods, the areas of secondary hydrogenating of cladding are spaced from the primary defects by ∼2500-3000 mm, as a rule, and are often adjacent closely to the upper welded joints. There is no pronounced dependence of the distance between the primary and secondary cladding defects neither on the linear power, at which the fuel rods were operated, nor on the period of their operation in the leaky state. The time period of the significant secondary damage formation is about 250 ± 50 calendar days for the WWER fuel rods with slight through primary defects (∼0.1 - 0.5 mm 2 ) operated in the linear power range 170-215 W/cm. Cladding degradation, taking place due to the secondary hydrogenating, does not occur in case of large through debris-defects during operation up to 600 calendar days

  6. Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

    International Nuclear Information System (INIS)

    Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

    1979-01-01

    Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

  7. 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.

  8. Influence of pellet-clad-gap-size on LWR fuel rod performance

    International Nuclear Information System (INIS)

    Brzoska, B.; Fuchs, H.P.; Garzarolli, F.; Manzel, R.

    1979-01-01

    The as-fabricated pellet-clad-gap size varies due to fabricational tolerances of the cladding inner diameter and the pellet outer diameter. The consequences of these variations on the fuel rod behaviour are analyzed using the KWU fuel rod code CARO. The code predictions are compared with experimental results of special pathfinder test fuel rods irradiated in the OBRIGHEIM nuclear power plant. These test fuel rods include gap sizer in the range of 140 μm to 270 μm, prepressurization between 13 bar to 36 bar and Helium and Argon fill gases irradiated up to a local burnup of 35 MWd/kg(U). Post irradiation examination were performed at different burnups. CARC calculations have been performed with special emphasis in cladding creep down, fission gas release and pellet clad gap closure. (orig.)

  9. Fuel rod-grid interaction wear: in-reactor tests (LWBR development program)

    International Nuclear Information System (INIS)

    Stackhouse, R.M.

    1979-11-01

    Wear of the Zircaloy cladding of LWBR irradiation test fuel rods, resulting from relative motion between rod and rod support contacts, is reported. Measured wear depths were small, 0.0 to 2.7 mils, but are important in fuel element behavior assessment because of the local loss of cladding thickness, as well as the effect on grid spring forces that laterally restrain the rods. An empirical wear analysis model, based on out-of-pile tests, is presented. The model was used to calculate the wear on the irradiation test fuel rods attributed to a combination of up-and-down motions resulting from power and pressure/temperature cycling of the test reactor, flow-induced vibrations, and assembly handling scratches. The calculated depths are generally deeper than the measured depths

  10. Spring retainer apparatus and method for facilitating loading of fuel rods into a fuel assembly grid

    International Nuclear Information System (INIS)

    De Mario, E.E.

    1988-01-01

    For use with a fuel assembly having at least one grid formed of interleaved straps defining hollow cells for respectively receiving fuel rods, at least some of the straps being disposed in pairs thereof so as to form springs in pairs therof being positioned in back-to-back relationships between adjacent ones of the cells, the springs in each pair thereof being configured to normally assume expanded positions in which they are displaced away from one another to engage fuel rods received in the respective cells and being deflectible to retracted positions in which they are displaced toward one another to allow loading of the fuel rods in the respective cells without engaging the springs, a spring retainer apparatus for facilitating the loading of the fuel rods into the cells of the fuel assembly grid is described comprising: (a) elongated holder bars, each holder bar being alignable with one of the pairs of the straps of the grid which defines the pairs of springs and extendible along, and in spaced relation from, the one strap pair and between and spaced from positions occupied by fuel rods when received in the cells of the grid; and (b) supported by each of the holder bars corresponding to the pairs of springs defined by the pair of straps aligned with the holder bar

  11. An Evaluation on the Fluid Elastic Instability of the Fuel Rod for OPR1000 Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeong Koo; Jeon, Sang Yoon; Lee, Kyu Seok; Kim, Jeong Ha; Lee, Sang Jong [Reactor Core Technology Department, Korea Nuclear Fuel, 493, Deogjin, Yuseong, Daejeon, 305-353 (Korea, Republic of)

    2009-06-15

    The fuel assembly for a typical PWR (Pressurized Water Reactor) plant suffers severe operating conditions during its lifetime such as high temperature, high pressure and massive coolant passing through the fuel assembly with high speed. Moreover, recently nuclear fuel is requested not only to operate under more severe operation conditions for example high burnup, longer cycle and power up-rate, but also to maintain its integrity in spite of the operation severity. Lots of vendors, therefore, have poured their endeavor to develop an advanced fuel in order to meet these requirements. However, the fuel failures are still reported from time to time. In general, fuel failure mechanisms known as significant causes of PWR fuel failure are grid to rod fretting, corrosion of the cladding, pellet cladding interaction and debris induced fretting. Especially, since the fuel assembly is very tall and flexible structure and the flow velocity of reactor coolant is pretty high, flow induced vibration (FIV) of fuel rod is an inevitable phenomenon in PWR fuel and the energy vibrating fuel rod continually provided by coolant flow can become a root cause of the fuel failure like grid to rod fretting. Moreover, the cross flow of the coolant is highly susceptible to cause the fluid elastic instability (FEI) which produces extraordinarily big amplitudes of the fuel rod suddenly and is eventually ended up fuel failure within very short-term. The FIV problem, therefore, has to be evaluated carefully to avoid unexpected fuel failure. At present, the susceptibility to vibration damage of the fuel rod for OPR1000 plants has been estimated by the comparison of natural frequencies of every fuel rod span with recognized external excitation frequencies like coolant pump blade passing frequencies, vortex shedding frequencies and lower support structure vibration frequencies. That is, in order to prevent fuel failure due to the external excitation, the natural frequencies of unsupported lengths of

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

    International Nuclear Information System (INIS)

    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

  13. Methods for acquiring data in power ramping experiments with WWER fuel rods at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Bobrov, S N; Grachev, A F; Ovchinnikov, V A; Poliakov, I S; Matveev, N P [Research Inst. of Atomic Reactors, Dimitrovgrad (Russian Federation); Novikov, V V [Institute of Inorganic Materials, Moscow (Russian Federation)

    1997-08-01

    A programme on in-pile test which involve fuel burnup up to 60 MWd/kg and up to 12 fuel rods in the experimental rig is considered. Testing methods with reference to the MIR-M1 reactor are reported. Power ramping regime can be realized either by an increase of the total reactor capacity or by displacement of the nearest to the experimental cell control rods or by combination of these two ways. A total thermal capacity of the fuel rod cluster is determined by means of the thermal balance technique. The thermal capacity of each separate fuel rod can be estimated from the distribution of their relative activity within the accuracy range 5-10%. The important condition for this procedure is to keep the initial distribution of the fuel rod heating during power ramping. Means of instrumentation are described. They are standard detectors of loop facilities and transducers installed both in the irradiation rigs and fuel rods. Different ways of processing data on the fuel rod loss of integrity are reported. When the time of fuel rod loss of tightness is placed in correspondence with its capacity, processing can be made either on the maximum fuel rod heat load or on that at crack location. The information acquired in the experiments on the burnup values, heat rating distribution, kinetics of fission product gas emission, fuel rod elongation, fuel rod diameter changes, crack availability and fission products migration is used for the development and verification of calculation codes. (author). 1 ref., 4 figs, 1 tab.

  14. Radial power density distribution of MOX fuel rods in the HBWR

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Joo, Hyung Kook; Lee, Byung Ho; Sohn, Dong Seong

    1999-07-01

    Two MOX fuel rods, which ar being fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with the Korea Atomic Energy Research Institute (KAERI), are going to be irradiated in the HBWR (Halden Boiling Water Reactor) from the beginning of 2000 in the framework of OECD Halden Reactor Programme (HRP) together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is a basic property in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR H BWR that calculates radial power density distribution for three MOX fuel rods have been developed subroutine FACTOR H BWR gives good agreement with the physics calculation except slight underprediction in the central part and a little overprediction at the outer part of the pellet. 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. (author). 5 refs., 3 tabs., 24 figs

  15. Fuel-rod response during the large-break LOCA Test LOC-6

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% 235 U). Each rod was surrounded by an individual flow shroud

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

    International Nuclear Information System (INIS)

    Caruso, S.

    2007-01-01

    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 235 U enrichment of the fresh UO 2 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 UO 2 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

  17. Apparatus for injection casting metallic nuclear energy fuel rods

    Science.gov (United States)

    Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

    1991-01-01

    Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

  18. Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

    International Nuclear Information System (INIS)

    Waseem; Elahi, N.; Siddiqui, A.; Murtaza, G.

    2011-01-01

    Research highlights: → A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. → The spring hold-down force is calculated using the contact pressure obtained from the FE model. → Experiment has also been conducted in the same environment for the measurement of this force. → The spring hold-down force values obtained from both studies confirm the validation of this analysis. → The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

  19. Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

    Energy Technology Data Exchange (ETDEWEB)

    Waseem, E-mail: wazim_me@hotmail.co [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan); Elahi, N.; Siddiqui, A.; Murtaza, G. [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan)

    2011-01-15

    Research highlights: A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies confirm the validation of this analysis. The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

  20. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test 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 technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  1. Nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Busch, H.; Mindnich, F.R.

    1973-01-01

    The fuel rod consists of a can with at least one end cap and a plenum spring between this cap and the fuel. To prevent the hazard that a eutectic mixture is formed during welding of the end cap, a thermal insulation is added between the end cap and plenum spring. It consists of a comical extension of the end cap with a terminal disc against which the spring is supported. The end cap, the extension, and the disc may be formed by one or several pieces. If the disc is separated from the other parts it may be manufactured from chrome steel or VA steel. (DG) [de

  2. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  3. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  4. Device for replacing the rods of a fuel element of a nuclear reactor

    International Nuclear Information System (INIS)

    Nissel, B.; Kybranz, R.; Will, R.

    1977-01-01

    In order to be able to replace several separate rods (fuel rods or absorber rods), in a fuel element, a special grab is introduced, which consists of several individual gripping devices and is operated by spring loading. (TK) [de

  5. Design characteristics of metallic fuel rod on its in-LMR performance

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang Hee Young; Nam, Cheol; Kim, Jong Oh

    1997-01-01

    Fuel design is a key feature to assure LMR safety goals. To date, a large effort had been devoted to develop metallic fuels at ANL's experimental breeder reactor (EBR-II). The major design and performance parameters investigated include; thermal conductivity and temperature profile; smear density; axial plenum; FCMI and cladding deformation including creep, and fission gas release. In order to evaluate the sensitivity of each parameter, in-LMR performances of metallic fuels are not only reviewed by the experiment results in literatures, but also key design characteristics according to the variation of metallic fuel rod design parameters are analyzed by using the MACSIS code which simulates in-reactor behaviors of metal fuel rod. In this study, key design characteristics and the criteria which must be considered to design fuel rod in LMR, are proposed and discussed. (author). 14 refs., 4 figs

  6. CFD analysis of blockage length on a partially blocked fuel rod

    International Nuclear Information System (INIS)

    Scuro, Nikolas Lymberis; Andrade, Delvonei Alves de; Angelo, Gabriel; Angelo, Edvaldo

    2017-01-01

    In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod. (author)

  7. CFD analysis of blockage length on a partially blocked fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Scuro, Nikolas Lymberis; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Gabriel [Centro Universitário FEI (UNIFEI), São Paulo, SP (Brazil). Dept. de Engenharia Mecânica; Angelo, Edvaldo, E-mail: nikolas.scuro@gmail.com, E-mail: delvonei@ipen.br, E-mail: gangelo@fei.edu.br, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, São Paulo, SP (Brazil). Escola da Engenharia. Grupo de Simulação Numérica

    2017-07-01

    In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod. (author)

  8. AFA 2G and AFA 3G fuel rod performance analysis

    International Nuclear Information System (INIS)

    Lu Huaquan; Liu Tong; Jiao Yongjun; Pang Hua

    2002-01-01

    For 18-months fuel cycle strategy in GNPJVC DAYA BAY unit 1/2, by means of COCCINEL, the fuel rod performance for AFA 3G and AFA 2G in transition cycle is analyzed. The design criteria which should be respected in fuel rod design are included and the design methodology is introduced. All the criteria mentioned are verified and met

  9. Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

    International Nuclear Information System (INIS)

    Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

    1979-01-01

    Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

  10. The development of the fuel rod transient performance analysis code FTPAC

    International Nuclear Information System (INIS)

    Han Zhijie; Ji Songtao

    2014-01-01

    Fuel rod behavior, especially the integrity of cladding, played an important role in fuel safety research during reactor transient and hypothetical accidents conditions. In order to study fuel rod performance under transient accidents, FTPAC (Fuel Transient Performance Analysis Code) has been developed for simulating light water reactor fuel rod transient behavior when power or coolant boundary conditions are rapidly changing. It is composed of temperature, mechanical deformation, cladding oxidation and gas pressure model. The assessment was performed by comparing FTPAC code analysis result to experiments data and FRAPTRAN code calculations. Comparison shows that, the FTPAC gives reasonable agreement in temperature, deformation and gas pressure prediction. And the application of slip coefficient is more suitable for simulating the sliding between pellet and cladding when the gap is closed. (authors)

  11. Transient fuel rod behavior prediction with RODEX-3/SIERRA

    Energy Technology Data Exchange (ETDEWEB)

    Billaux, M R; Shann, S H; Swam, L.F. Van [Siemens Power Corp., Richland, WA (United States)

    1997-08-01

    This paper discusses some aspects of the fuel performance code SIERRA (SIEmens Rod Response Analysis). SIERRA, the latest version of the code RODEX-3, has been developed to improve the fuel performance prediction capabilities of the code, both at high burnup and during transient reactor conditions. The paper emphasizes the importance of the mechanical models of the cracked pellet and of the cladding, in the prediction of the transient response of the fuel rod to power changes. These models are discussed in detail. Other aspects of the modelling of high burnup effects are also presented, in particular the modelling of the rim effect and the way it affects the fuel temperature. (author). 12 refs, 5 figs.

  12. Transient fuel rod behavior prediction with RODEX-3/SIERRA

    International Nuclear Information System (INIS)

    Billaux, M.R.; Shann, S.H.; Swam, L.F. Van

    1997-01-01

    This paper discusses some aspects of the fuel performance code SIERRA (SIEmens Rod Response Analysis). SIERRA, the latest version of the code RODEX-3, has been developed to improve the fuel performance prediction capabilities of the code, both at high burnup and during transient reactor conditions. The paper emphasizes the importance of the mechanical models of the cracked pellet and of the cladding, in the prediction of the transient response of the fuel rod to power changes. These models are discussed in detail. Other aspects of the modelling of high burnup effects are also presented, in particular the modelling of the rim effect and the way it affects the fuel temperature. (author). 12 refs, 5 figs

  13. Stress Analysis of Single Spacer Grid Support considering Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Y. G.; Jung, D. H.; Kim, J. H. [Chungnam National University, Daejeon (Korea, Republic of); Park, J. K.; Jeon, K. L. [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2010-10-15

    Pressurized water reactor (PWR) nuclear fuel assembly is mainly composed of a top-end piece, a bottom-end piece, lots of fuel rods, and several spacer grids. Among them, the main function of spacer grid is protecting fuel rods from Fluid Induced Vibration (FIV). The cross section of spacer grid assembled by laser welding in upper and lower point. When the fuel rod inserted in spacer gird, spring and dimple and around of welded area got a stresses. The main hypothesis of this analysis is the boundary area of HAZ and base metal can get a lot of damage than other area by FIV. So, design factors of spacer grid mainly considered to preventing the fatigue failure in HAZ and spring and dimple of spacer grid. From previous researching, the environment in reactor verified. Pressure and temperature of light water observed 15MPa and 320 .deg. C, and vibration of the fuel rod observed within 0 {approx} 50Hz. In this study, mechanical properties of zirconium alloy that extracted from the test and the spacer grid model which used in the PWR were applied in stress analyzing. General-purpose finite element analysis program was used ANSYS Workbench 12.0.1 version. 3-D CAD program CATIA was used to create spacer grid model

  14. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  15. Apparatus for inspecting the quality of nuclear fuel rod ends

    International Nuclear Information System (INIS)

    Brashier, R.W.; Pfau, E.D.

    1990-01-01

    This patent describes an apparatus for inspecting the quality of both ends of nuclear fuel rods. It comprises: a housing including a pair of longitudinally separated slots for receiving X-ray downwardly therethrough from an external source and so as to define first and second longitudinally spaced apart operating positions, means for serially guiding nuclear fuel rods longitudinally through the housing and to a first rod position wherein the forward ends of the rods are aligned below the first operating position and to a second rod position wherein the rear ends of the rods are aligned below the second operating position, belt conveyor assembly means for serially advancing X-ray film cartridges longitudinally through the housing and below the rods, and so that each cartridge may be selectively aligned below the first and second operating positions; and table means supported by the conveyor frame for selectively lifting the film cartridges supported by the belts and so that the conveyor belts may be advanced while the film cartridges are held stationary

  16. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Andrade, G.G. de

    1982-01-01

    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.) [pt

  17. Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

  18. Analysis of Radial Plutonium Isotope Distribution in Irradiated Test MOX Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jae Yong; Lee, Byung Ho; Koo, Yang Hyun; Kim, Han Soo

    2009-01-15

    After Rod 3 and 6 (KAERI MOX) were irradiated in the Halden reactor, their post-irradiation examinations are being carried out now. In this report, PLUTON code was implemented to analyze Rod 3 and 6 (KAERI MOX). In the both rods, the ratio of a maximum burnup to an average burnup in the radial distribution was 1.3 and the contents of {sup 239}Pu tended to increase as the radial position approached the periphery of the fuel pellet. The detailed radial distribution of {sup 239}Pu and {sup 240}Pu, however, were somewhat different. To find the reason for this difference, the contents of Pu isotopes were investigated as the burnup increased. The content of {sup 239}Pu decreased with the burnup. The content of {sup 240}Pu increased with the burnup by the 20 GWd/tM but decreased over the 20 GWd/tM. The local burnup of Rod 3 is higher than that of Rod 6 due to the hole penetrated through the fuel rod. The content of {sup 239}Pu decreased more rapidly than that of {sup 240}Pu in the Rod 6 with the increased burnup. It resulted in a radial distribution of {sup 239}Pu and {sup 240}Pu similar to Rod 3. The ratio of Xe to Kr is a parameter to find where the fissions occur in the nuclear fuel. In both Rod 3 and 6, it was 18.3 in the whole fuel rod cross section, which showed that the fissions occurred in the plutonium.

  19. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  20. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Science.gov (United States)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  1. 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)

  2. A user input manual for single fuel rod behaviour analysis code FEMAXI-III

    International Nuclear Information System (INIS)

    Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

    1983-03-01

    Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

  3. Multidimensional simulations of hydrides during fuel rod lifecycle

    International Nuclear Information System (INIS)

    Stafford, D.S.

    2015-01-01

    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. - Highlights: • We extend BISON fuel performance code to simulate lifecycle of fuel rods. • We model hydrogen evolution in cladding from reactor through dry storage. • We validate 1D simulations of hydrogen evolution against experiments. • We show results of 2D axisymmetric simulations predicting hydride formation. • We show how our model predicts formation of a hydride rim in the cladding.

  4. Secondary hydriding of defected zircaloy-clad fuel rods

    International Nuclear Information System (INIS)

    Olander, D.R.; Vaknin, S.

    1993-01-01

    The phenomenon of secondary hydriding in LWR fuel rods is critically reviewed. The current understanding of the process is summarized with emphasis on the sources of hydrogen in the rod provided by chemical reaction of water (steam) introduced via a primary defect in the cladding. As often noted in the literature, the role of hydrogen peroxide produced by steam radiolysis is to provide sources of hydrogen by cladding and fuel oxidation that are absent without fission-fragment irradiation of the gas. Quantitative description of the evolution of the chemical state inside the fuel rod is achieved by combining the chemical kinetics of the reactions between the gas and the fuel and cladding with the transport by diffusion of components of the gas in the gap. The chemistry-gas transport model provides the framework into which therate constants of the reactions between the gases in the gap and the fuel and cladding are incorporated. The output of the model calculation is the H 2 0/H 2 ratio in the gas and the degree of claddingand fuel oxidation as functions of distance from the primary defect. This output, when combined with a criterion for the onset of massive hydriding of the cladding, can provide a prediction of the time and location of a potential secondary hydriding failure. The chemistry-gas transport model is the starting point for mechanical and H-in-Zr migration analyses intended to determine the nature of the cladding failure caused by the development of the massive hydride on the inner wall

  5. Thermal-stress analysis of HTGR fuel and control rod fuel blocks in in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    The equivalent solid plate method, in conjunction with two-dimensional plane stress and plane strain analyses, was used in assessing the thermal stress behavior of HTGR fuel and control rod fuel blocks. For the control rod fuel blocks, particular attention was given to ascertaining the effects of the reserve shutdown hole and the control rod channel holes. The assumed safety factor of 2 on the failure criteria was considered adequate to account for neglecting the axial temperature gradient in the plane analyses of the ends of the blocks. The analyses indicated that the maximum calculated tensile stress values were smaller than the criteria values except for the plane strain analysis of the control rod fuel block end surfaces and the axisymmetric analysis of the fuel block as a circular cylinder. However, most of the maximum calculated strain values were greater than the criteria values

  6. Determination and microscopic study of incipient defects in irradiated power reactor fuel rods. Final report

    International Nuclear Information System (INIS)

    Pasupathi, V.; Perrin, J.S.; Roberts, E.

    1978-05-01

    This report presents the results of nondestructive and destructive examinations carried out on the Point Beach-1 (PWR) and Dresden-3 (BWR) candidate fuel rods selected for the study of pellet-clad interaction (PCI) induced incipient defects. In addition, the report includes results of examination of sections from Oskarshamn-1 (BWR) fuel rods. Eddy current examination of Point Beach-1 rods showed indications of possible incipient defects in the fuel rods. The profilometry and the gamma scan data also indicated that the source of the eddy current indications may be incipient defects. No failed rods or rods with incipient failure were found in the sample from Point Beach-1. Despite the lack of success in finding incipient defects and filed rods, the mechanism for fuel rod failures in Point Beach-1 is postulated to be PCI-related, with high startup rates and fuel handling being the key elements. Nine out of the 10 candidate fuel rods from Dresden-3 (BWR) were failed, and all the failed rods had leaked water so that the initial mechanism was observed. Examination of clad inner surfaces of the specimens from failed and unfailed rods showed fuel deposits of widely varying appearance. The deposits were found to contain uranium, cesium, and tellurium. Transmission electron microscopy of clad specimens showed evidence of microscopic strain. Metallographic examination of fuel pellets from the peak transient power location showed extensive grain boundary separation and axial movement of the fuel indicative of rapid release of fission products. Examination of Oskarshamn clad specimens did not show any stress corrosion crack (SCC) type defects. The defects found in the examinations appear to be related to secondary hydriding. The clad inner surface of the Oskarshamn specimens also showed uranium-rich deposits of varying features

  7. Irradiation of pressurized water reactor fuel rods in the Forschungsreaktor Juelich 2

    International Nuclear Information System (INIS)

    Gaertner, M.

    1978-10-01

    Test fuel rods have been irradiated in FRJ-2 to study the interaction between fuel and cladding as well as hydride orientation stability in the prehydrided cladding. The fuel rods achieved burn-ups of 3.500 to 10.000 MWd/tU at surface temperatures of 333 0 C and power levels up to 620 W/cm. (orig.) [de

  8. HTGR fuel rods: carbon-carbon composites designed for high weight and low strength

    International Nuclear Information System (INIS)

    Bullock, R.E.

    1977-01-01

    The evolution of the process for fabricating fuel rods for the high-temperature gas-cooled reactor (HTGR) by injection and carbonization of a thermoplastic matrix that bonds close-packed beds of pyrocarbon-coated fuel particles together is reviewed for the fresh-fuel cycle, and a variant process involving a thermosetting matrix that would allow free-standing carbonization of refabricated fuel is discussed. Previous attempts to fabricate such injection-bonded fuel rods from undiluted thermosetting binders filled with powdered graphite were unsuccessful, because of damage to coatings on fuel particles that resulted from strong particle-to-matrix bonding in conjunction with large matrix shrinkage on carbonization and subsequent irradiation. These problems have now been overcome through the use of a diluted thermosetting matrix with a low-char-yield additive (fugitive), which produces a more porous char similar to that from the pitch-based thermoplastic used in fabrication of fresh fuel. A 1-to-1 dilution of resin with fugitive produced the optimum binder for injection and carbonization, where the fired matrix in such rods contained about 20 wt% binder char and 80 wt% powdered graphite. Thermosetting fuel rods diluted with various amounts of fugitive to give binder chars that range from 12 to 48 wt% of the fired matrix have been subjected to irradiation screening tests, and rods with no more than 32 wt% binder char appear to perform about as well under irradiation as do pitch-based rods. However, particle damage does begin to occur in those lightly diluted rods in which the less-stable binder char constitutes more than 32 wt% of the fired matrix. (author)

  9. Critical heat flux tests for self-spaced square finned 7 fuel rod bundle

    International Nuclear Information System (INIS)

    Moon, Sang Ki; Chun, Se Young; Choi, Ki Young; Park, Jong Kuk; Hwang, Dae Hyun; Zee, Sung Quun; Kim, Keung Koo

    2001-09-01

    Now, KAERI is developing a new advanced reactor aimed at achieving highly enhanced safety and reliability, and improved economics. SSF (Self-Spaced Square Finned) fuel rod bundle is considered as a suitable one for the new advanced reactor. The SSF fuel rods have rectangular shapes and four fins at the corners, and are arranged in triangular geometry. While the SSF fuel rod bundle is considered to have enhanced cooling efficiency, the correlations used for commercial PWR might be able to be applied. The application results of some conventional correlations show that the SSF fuel rod bundle show an enhanced CHF performance about 10 to 40 %. When some conventional CHF correlations are applied to CHF data with a similar geometry to the SSF fuel rod bundle, conventional CHF correlations including a correlation developed in Russia are judged not to be suitable for the development of SSF fuel rod bundle and for the use in a safety analysis code. From CHF experiments for SSF 7 fuel rod bundle performed in KAERI, the following results are obtained: the CHF increases with increasing mass flux, and the CHF increasing rate decreases at high mass flux conditions. The exit quality decreases with increasing mass flux. The overall effect of the mass flux on the CHF and exit quality coincides with previous understanding. Compared to the CHF data of IPPE with the same system pressure and inlet temperature, the CHF data of KAERI show the similar values. Thus, the reliability of IPPE CHF data can be confirmed indirectly

  10. Corrosion behaviour of zircaloy 4 fuel rod cladding in EDF power plants

    Energy Technology Data Exchange (ETDEWEB)

    Romary, H; Deydier, D [EDF, Direction de l` Equipment SEPTEN, Villeurbanne (France)

    1997-02-01

    Since the beginning of the French nuclear program, a surveillance of fuel has been carried out in order to evaluate the fuel behaviour under irradiation. Until now, nuclear fuels provided by suppliers have met EDF requirements concerning fuel behaviour and reliability. But, the need to minimize the costs and to increase the flexibility of the power plants led EDF to the definition of new targets: optimization of the core management and fuel cycle economy. The fuel behaviour experience shows that some of these new requirements cannot be fully fulfilled by the present standard fuel due to some technological limits. Particularly, burnup enhancement is limited by the oxidation and the hydriding of the Zircaloy 4 fuel rod cladding. Also, fuel suppliers and EDF need to have a better knowledge of the Zy-4 cladding behaviour in order to define the existing margins and the limiting factors. For this reason, in-reactor fuel characterization programs have been set up by fuel suppliers and EDF for a few years. This paper presents the main results and conclusions of EDF experience on Zy-4 in-reactor corrosion behaviour. Data obtained from oxide layer or zirconia thickness measurements show that corrosion performance of Zy-4 fuel rod cladding, as irradiated until now in EDF reactors, is satisfactory but not sufficient to meet the future needs. The fuel suppliers propose in order to improve the corrosion resistance of fuel rod cladding, low tin Zy-4 cladding and then optimized Zy-4 cladding. Irradiation of these claddings are ongoing. The available corrosion data show the better in-reactor corrosion resistance of optimized Zy-4 fuel rod cladding compared to the standard Zy-4 cladding. The scheduled fuel surveillance program will confirm if the optimized Zy-4 fuel rod cladding will meet the requirements for the future high burnup and high flexibility fuel. (author). 10 refs, 19 figs, 4 tabs.

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

    International Nuclear Information System (INIS)

    Toebbe, H.

    1990-05-01

    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

  12. Lateral Flow Field Behavior Downstream of Mixing Vanes In a Simulated Nuclear Fuel Rod Bundle

    International Nuclear Information System (INIS)

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2004-01-01

    To assess the fuel assembly performance of PWR nuclear fuel assemblies, average subchannel flow values are used in design analyses. However, for this highly complex flow, it is known that local conditions around fuel rods vary dependent upon the location of the fuel rod in the fuel assembly and upon the support grid design that maintains the fuel rod pitch. To investigate the local flow in a simulated nuclear fuel rod bundle, a testing technique has been employed to measure the lateral flow field in a 5 x 5 rod bundle. Particle Image Velocimetry was used to measure the lateral flow field downstream of a support grid with mixing vanes for four unique subchannels in the 5 x 5 bundle. The dominant lateral flow structures for each subchannel are compared in this paper including the decay of these flow structures. (authors)

  13. Thermal phenomenae in nuclear fuel rods

    International Nuclear Information System (INIS)

    Baigorria, Carlos.

    1983-12-01

    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) [es

  14. Development of thermocouple re-instrumentation technique for irradiated fuel rod. Techniques for making center hole into UO2 pellets and thermocouple re-instrumentation to fuel rod

    International Nuclear Information System (INIS)

    Shimizu, Michio; Saito, Junichi; Oshima, Kunio

    1995-07-01

    The information on FP gas pressure and centerline temperature of fuel pellets during power transient is important to study the pellet clad interaction (PCI) mechanism of high burnup LWR fuel rods. At the Department of JMTR, a re-instrumentation technique of FP gas pressure gage for an irradiated fuel rod was developed in 1990. Furthermore, a thermocouple re-instrumentation technique was successfully developed in 1994. Two steps were taken to carry out the development program of the thermocouple re-instrumentation technique. In the first step, a drilling technique was developed for making a center hole of the irradiated fuel pellets. Various drilling tests were carried out using dummy of fuel rods consisted of Ba 2 FeO 3 pellets and Zry-2 cladding. On this work it is important to keep the pellets just the state cracked at a power reactor. In these tests, the technique to fix the pellets by frozen CO 2 was used during the drilling work. Also, diamond drills were used to make the center hole. These tests were completed successfully. A center hole, 54mm depth and 2.5mm diameter, was realized by these methods. The second step of this program is the in-pile demonstration test on an irradiated fuel rod instrumented dually a thermocouple and FP gas pressure gage. The demonstration test was carried out at the JMTR in 1995. (author)

  15. Vibration mechanism of fuel rod in axial flow

    International Nuclear Information System (INIS)

    Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Song, Kee Nam

    1998-08-01

    This is a review on the previous researches for the vibration of fuel rod induced by axial flow. The analysis methods are classified into three categories accordingly as the researchers postulate the vibration to be self-excited, forced and parametric; the self-excited mechanism by Burgreen and Quinn, the forced one by Reavis, Gorman, kanazawa, and S. Chen, and the parametric one by Y. Chen. Quinn supposed that the centrifugal force by flow exaggerated the natural bow in the cylinder, and the flexural force by it diminished the bow by turns; this interactive motion leaded cylinder to vibration. The supporters to the forced mechanism considered the forces arising from pressure perturbation within the boundary layers as vibrating sources. Y. Chen insisted that the cylinder could only be excited to vibration in resonance by the small oscillation of mean flow velocity. The previous studies were based on the simple boundary conditions such as hinged-hinged or fixed-fixed single span. Therefore, for the more accurate prediction of the fuel rod vibration in reactor, the further studies need to reflect the actual boundary conditions of the fuel rod like axial force and continuous supports by grids. (author). 25 refs

  16. State of fuel rods spent in the VVER-1000 reactor up to a fuel burnup of 75 MW·Day/KgU

    International Nuclear Information System (INIS)

    Markov, D.; Zvir, E.; Polenok, V.; Zhitelev, V.; Strozhuk, A.; Volkova, I.

    2011-01-01

    The presented material contains the data on change in form, corrosion state and mechanical properties of fuel rod claddings, change in fuel structure and release of gaseous fission products (GFP) under the cladding. The results of PIEs of the VVER-1000 fuel rods with the high burnup of fuel (average value is 72.3 MW·day/kgU and maximum is 75 MW·day/kgU) carried out in JSC 'SSC RIAR' show that by the basic operational characteristics the lifetime of fuel rods with such burnup of fuel is not exhausted. The state of fuel rods is characterized by following key parameters. The fuel-to-cladding gap on the most part of the fuel meat is absent. With the burnup growth, diameter of the fuel rod increases due to fuel meat swelling. In so doing, the reverse strain achieves the values of 0.40-0.47 %. Ridges on the cladding are formed practically along the entire length of the fuel meat, average height of ridges makes up 25 μm, maximum - 40 μm. At burnups exceeding 55 MW·day/kgU, the rate of the fuel rod elongation is less than at low and average burnups. So if within a burnup range of 20-55 MW·day/kgU, the rate of the fuel rod elongation makes up about 0.330mm per 1 MW·day/kgU, at burnups exceeding 55 MW·day/kgU it is only 0.085mm per 1 MW·day/kgU. Corrosion state of the claddings of fuel rods with high burnup of fuel is satisfactory. The oxide film, as a rule, is uniform, dense, without cracks and exfoliation, its thickness on the external surface does not exceed 13 μm, while on the internal surface - 15 μm. Hydrogenation is insignificant, mass fraction of hydrogen does not exceed 0.01 %. Interaction of fuel rods with spacer grids does not result in significant fretting-corrosion. Based of the results of tests, short-term mechanical properties of the claddings of fuel rods with high burnup of fuel remain at high level. The state of fuel is characterized by absence of the fuel-to-cladding gap on the most part of the fuel meat, fuel is tightly fixed to the cladding

  17. Nondestructive examination of irradiated fuel rods by pulsed eddy current techniques

    International Nuclear Information System (INIS)

    Francis, W.C.; Quapp, W.J.; Martin, M.R.; Gibson, G.W.

    1976-02-01

    A number of fuel rods and unfueled zircaloy cladding tubes which had been irradiated in the Saxton reactor have undergone extensive nondestructive and corroborative destructive examinations by Aerojet Nuclear Company as part of the Water Reactor Safety Research Program, Irradiation Effects Test Series. This report discusses the pulsed eddy current (PEC) nondestructive examinations on the fuel rods and tubing and the metallography results on two fuel rods and one irradiated zircaloy tube. The PEC equipment, designed jointly by Argonne National Laboratory and Aerojet, performed very satisfactorily the functions of diameter, profile, and wall thickness measurements and OD and ID surface defect detection. The destructive examination provided reasonably good confirmation of ''defects'' detected in the nondestructive examination

  18. Generation of heat on fuel rod in cosine pattern by using induction heating

    International Nuclear Information System (INIS)

    Keettikkal, Felix; Sajeesh, Divya; Rao, Poornima; Hande, Shashank; Dakave, Ganesh; Kute, Tushar; Mahajan, Akshay; Kulkarni, R.D.

    2017-01-01

    Fuel rods are used in a nuclear reactor for fission process. When these rods are cooled by water during the heat transfer, the temperature stress causes undesirable defects in the fuel rod. Studying these defects occurring in the fuel rod in the nuclear cluster during nuclear reaction is a difficult task because fission reaction makes it difficult to analyse the changes in the rod. Hence there is a need to use a replica of the rod with similar thermal stress to study and analyse the rod for the defects. Normally the heat generated on the fuel rod follows a cosine pattern which is an inherent characteristic inside a nuclear reactor. In view of this, in this paper induction heating method is used on a rod to create an exact replica of the cosine pattern of heat by varying the pitch of the coil. First, a MATLAB simulation is done using simulink. Then a prototype of the model has been developed comprising of carbon steel pipe, with length and outside diameter of 1 meter and 48.2 mm, respectively. Instead of using water as coolant, rod is simulated in air. Therefore, the heat generated is lost by normal convection and radiation. Non-nuclear testing can be a valuable tool in the development or in some kind of experiment using nuclear reactor. Induction heating becomes an alternative to classical heating technologies because of its advantages such as efficiency, quickness, safety, clean heating and accurate power control. (author)

  19. Assessment of US NRC fuel rod behavior codes to extended burnup

    International Nuclear Information System (INIS)

    Laats, E.T.; Croucher, D.W.; Haggag, F.M.

    1982-01-01

    The purpose of this paper is to report the status of assessing the capabilities of the NRC fuel rod performance codes for calculating extended burnup rod behavior. As part of this effort, a large spectrum of fuel rod behavior phenomena was examined, and the phenomena deemed as being influential during extended burnup operation were identified. Then, the experiment data base addressing these identified phenomena was examined for availability and completeness at extended burnups. Calculational capabilities of the NRC's steady state FRAPCON-2 and transient FRAP-T6 fuel rod behavior codes were examined for each of the identified phenomenon. Parameters calculated by the codes were compared with the available data base, and judgments were made regarding model performance. Overall, the FRAPCON-2 code was found to be moderately well assessed to extended burnups, but the FRAP-T6 code cannot be adequately assessed until more transient high burnup data are available

  20. Change in geometrical parameters of WWER high burnup fuel rods under operational conditions and transient testing

    International Nuclear Information System (INIS)

    Kanashov, B.; Amosov, S.; Lyadov, G.; Markov, D.; Ovchinnikov, V; Polenok, V.; Smirnov, A.; Sukhikh, A.; Bek, E.; Yenin, A.; Novikov, V.

    2001-01-01

    The paper discusses changes in fuel rods geometric parameters as result of operation conditions and burnups. The degree of geometry variability of fuel rods, cladding and column is one of the most important characteristics affecting fuel serviceability. On the other hand, changes in fuel rod geometric parameters influence fuel temperature, fission gas release, fuel-to-cladding stress strained state as well as the degree of interaction with FA skeleton elements and skeleton rigidity. Change in fuel-to-cladding gap is measured using compression technique. The axial distribution of fuel-to-cladding gap demonstrates the largest decrease of the gap in the region 500 to 2000 mm from the bottom of the fuel rod (WWER-440) and in the region of 500 to 3000 mm for WWER-1000. The cladding material creep in WWER fuel rods together with the radiation growth results in fuel rod cladding elongation. A set of transient tests for spent WWER-440 and WWER-1000 fuel rods carried out in SSC RIAR during a period 1995-1999, with the aim to estimate the changes in geometric parameters of FRs. The estimation of changes in outer diameter of cladding and fuel column and fuel-to-cladding gap are performed in transient conditions (changes in linear power range of 180 to 400 W/cm) for both WWER-440 and WWER-1000. WWER-440 fuel rods having the same burnup and close fuel-cladding contact before testing are subjected to considerable hoop cladding strain in testing up to 300 W/cm. But the hoop strain does not grow due to the structural changes in fuel column and decrease in central hole diameter occurred when the power is higher

  1. Production and release of the fission gas in (Th U)O2 fuel rods

    International Nuclear Information System (INIS)

    Dias, Marcio S.

    1982-06-01

    The volume, composition and release of the fission gas products were caculated for (Th, U)O 2 fuel rods. The theorectical calculations were compared with experimental results available on the literature. In ThO 2 + 5% UO 2 fuel rods it will be produced approximated 5% more fission gas as compared to UO 2 fuel rods. The fission gas composition or Xe to Kr ratio has showed a decreasing fuel brunup dependence, in opposition to that of UO 2 . Under the same fuel rod operational conditions, the (Th, U)O 2 fission gas release will be smaller as compared to UO 2 . This behaviour of (Th, U)O 2 fuel comes from smallest gas atom difusivity and higher activation energies of the processes that increase the fission gas release. (Author) [pt

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

    International Nuclear Information System (INIS)

    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 0 F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs

  3. Fuel and control rod failure behavior during degraded core accidents

    International Nuclear Information System (INIS)

    Chung, K.S.

    1984-01-01

    As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

  4. Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

    International Nuclear Information System (INIS)

    Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

    2013-01-01

    The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

  5. Use of a commercial heat transfer code to predict horizontally oriented spent fuel rod temperatures

    International Nuclear Information System (INIS)

    Wix, S.D.; Koski, J.A.

    1992-01-01

    Radioactive spent fuel assemblies are a source of hazardous waste that will have to be dealt with in the near future. It is anticipated that the spent fuel assemblies will be transported to disposal sites in spent fuel transportation casks. In order to design a reliable and safe transportation cask, the maximum cladding temperature of the spent fuel rod arrays must be calculated. The maximum rod temperature is a limiting factor in the amount of spent fuel that can be loaded in a transportation cask. The scope of this work is to demonstrate that reasonable and conservative spent fuel rod temperature predictions can be made using commercially available thermal analysis codes. The demonstration is accomplished by a comparison between numerical temperature predictions, with a commercially available thermal analysis code, and experimental temperature data for electrical rod heaters simulating a horizontally oriented spent fuel rod bundle

  6. Test requirement for PIE of HANARO irradiated fuel rod

    International Nuclear Information System (INIS)

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

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U 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. Characteristics of axial splits in failed BWR fuel rods

    International Nuclear Information System (INIS)

    Lysell, G.; Grigoriev, V.

    2000-01-01

    Secondary cladding defects in BWR fuel sometimes have the shape of long axial cracks or ''splits''. Due to the large open UO 2 surfaces exposed to the water, fission product and UO 2 release to the coolant can reach excessive levels leading to forced shut downs to remove the failed fuel rods. A number of such fuel rods have been examined in Studsvik over the last 10 years. The paper describes observations from the PIE of long cracks and discusses the driving force of the cracks. Details such as starting cracks, macroscopic and microscopic fracture surface appearance, cross sections of cracks, hydride precipitates, location and degree of plastic deformation are given. (author)

  8. Non-parametric order statistics method applied to uncertainty propagation in fuel rod calculations

    International Nuclear Information System (INIS)

    Arimescu, V.E.; Heins, L.

    2001-01-01

    Advances in modeling fuel rod behavior and accumulations of adequate experimental data have made possible the introduction of quantitative methods to estimate the uncertainty of predictions made with best-estimate fuel rod codes. The uncertainty range of the input variables is characterized by a truncated distribution which is typically a normal, lognormal, or uniform distribution. While the distribution for fabrication parameters is defined to cover the design or fabrication tolerances, the distribution of modeling parameters is inferred from the experimental database consisting of separate effects tests and global tests. The final step of the methodology uses a Monte Carlo type of random sampling of all relevant input variables and performs best-estimate code calculations to propagate these uncertainties in order to evaluate the uncertainty range of outputs of interest for design analysis, such as internal rod pressure and fuel centerline temperature. The statistical method underlying this Monte Carlo sampling is non-parametric order statistics, which is perfectly suited to evaluate quantiles of populations with unknown distribution. The application of this method is straightforward in the case of one single fuel rod, when a 95/95 statement is applicable: 'with a probability of 95% and confidence level of 95% the values of output of interest are below a certain value'. Therefore, the 0.95-quantile is estimated for the distribution of all possible values of one fuel rod with a statistical confidence of 95%. On the other hand, a more elaborate procedure is required if all the fuel rods in the core are being analyzed. In this case, the aim is to evaluate the following global statement: with 95% confidence level, the expected number of fuel rods which are not exceeding a certain value is all the fuel rods in the core except only a few fuel rods. In both cases, the thresholds determined by the analysis should be below the safety acceptable design limit. An indirect

  9. Method and apparatus for the production of a nuclear fuel rod

    International Nuclear Information System (INIS)

    Ballard, A.S.; Cooper, R.G.; Davis, D.E.

    1975-01-01

    The method designs the manufacture of e.g. rod-shaped fuel element fillings in which fuel particles are suspended within a liquid and solidifiable binder such as graphite powder in pitch. The fuel particles are filled into cavities whose cross-sections correspond to those of the fuel rods. After closing with a covering plate, a piston exerts a force from below on it until its solidification. To follow, the liquid binder is injected through lower openings in the cavities. Due to the lubricity of the binder, the cavities are heated to 150 to 175 0 C, the packing of particles are homogenized. This procedure is further supported by the constant pressure of the pistons. Excess binder and air can flow out through openings in the covering plate. After cooling and solidification of the binder as well as after removal of the covering plate, the piston thrusts out the formed bodies or fuel rods from the cavities by an upwards movement. (DG/LH) [de

  10. Preliminary Study on the Fretting Wear Behaviors of a Duel Cooled Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y.H.; Lee, K.H.; Kim, H.K. [KAERI, 150 Dukjin-dong Yuseon-gu Daejeon, 305-353 (Korea, Republic of)

    2009-06-15

    Based on MIT's concept, an innovative fuel development project was launched by KAERI that a substantial power up-rating could be realized by introducing an internally and externally double cooled annular fuel for current PWR reactors. In order to apply this duel cooled fuel to an OPR 1000 reactor system, geometrical features of structural parts in a fuel assembly should be changed except an overall dimension of a fuel assembly. Typical changes are summarized as fuel rod diameter and weight, shape and position of a spacer grid spring, etc. When considering a duel cooled fuel rod, its vibration characteristic and fretting behavior should be verified because the modified shape and dimension of spacer grid spring, fuel rod diameter and weight, number of spacer grid assembly are closely related to a flow-induced vibration in a duel cooled fuel assembly. In this study, based on FIV test results of 4x4 fuel assembly, fretting wear tests of an outer duel cooled fuel rod were performed by using an embossing type spacer grid spring that could adjust its spring stiffness. The discussion was focused on the evaluation of the optimized spring stiffness and spring position in 1x1 cell by analyzing the fretting wear results. (authors)

  11. Development of a program for evaluating the temperature of SMART-P fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Cheon, Jin Sik; Lee, Byung Ho; Koo, Yang Hyun; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong

    2003-11-01

    A code for evaluating the temperature of SMART-P fuel rod has been developed. Finite Element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for SMART-P fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. Also, given a power depression in fuel meat as a function of burnup, its effect on the centerline temperature was more precisely evaluated by the developed program compared to the ABAQUS code. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes.

  12. BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

    Energy Technology Data Exchange (ETDEWEB)

    Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sweet, Ryan T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling. As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations

  13. Detection of defective fuel rods in water reactors - a review

    International Nuclear Information System (INIS)

    Hartog, J.M.

    1980-01-01

    Consideration of the fundamental processes of fission product release within fuel pellets and at the pellet surface, and its transport in the fuel/cladding interspace and from fuel rod to coolant, indicates what radio-nuclides will be detectable in the coolant from small and large cladding failures. A better understanding of the aggregate fission product transport is required to allow reactor operators to interpret signals from detection systems in terms of quantitative cladding deterioration. This needs experimental investigation in a specially instrumented loop, as well as development of a technique to cause a rod to defect deliberately during steady power operation. (author)

  14. Control rod

    International Nuclear Information System (INIS)

    Kawakami, Kazuo; Shimoshige, Takanori; Nishimura, Akira

    1979-01-01

    Purpose: A control rod has been developed, which provided a plurality of through-holes in the vicinity of the sheath fitting position, in order to flatten burn-up, of fuel rods in positions confronting a control rod. Thereby to facilitate the manufacture of the control rods and prevent fuel rod failures. Constitution: A plurality of through-holes are formed in the vicinity of the sheath fitting position of a central support rod to which a sheath for the control rod is fitted. These through-holes are arranged in the axial direction of the central support rod. Accordingly, burn-up of fuel rods confronting the control rods can be reduced by through-holes and fuel rod failures can be prevented. (Yoshino, Y.)

  15. In-pile experiments on fuel rod behaviour during a LOCA

    International Nuclear Information System (INIS)

    Sepold, E.H.; Karb, E.H.; Pruessmann, M.

    1981-07-01

    This report describes the results of the Test Series G2/3 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program ist the burnup, ranging from 2500 to 35000 MWd/t. The results of test series G2/3 (35000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  16. In-pile experiemts on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Pruessmann, M.; Karb, E.H.; Sepold, L.

    1981-02-01

    This report describes the results of the Test Series G1 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechansims of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program is the burnup ranging from 2500 to 35 000 MWd/t. The results of test series G1 (35 000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  17. On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

    Science.gov (United States)

    Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

    2016-08-01

    For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. Copyright © 2016 Elsevier Ltd. All rights reserved.

  18. Probabilistic analysis of strength and thermal-physic WWER fuel rod characteristics using START-3 code

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Khramtsov; Sokolov, F.

    2001-01-01

    During the last years probabilistic methods for evaluation of the influence of the fuel geometry and technology parameters on fuel operational reliability are widely used. In the present work the START-3 procedure is used to calculate the thermal physics and strength characteristics of WWER fuel rods behavior. The procedure is based on the Monte-Carlo method with the application of Sobol quasi-random sequences. This technique allows to treat the fuel rod technological and operating parameters as well as its strength and thermal physics characteristics as random variables. The work deals with a series of WWER-1000 fuel rod statistical tests and verification based on the PIE results. Also preliminary calculations are implemented with the aim to determine the design schema parameters. This should ensure the accuracy of the assessment of the parameters of WWER fuel rod characteristics distribution. The probability characteristics of fuel rod strength and thermal physics are assessed via the statistical analysis of the results of probability calculations

  19. Nondestructive assay of spent fuel rods from a Light Water Breeder Reactor (LWBR Development Program)

    International Nuclear Information System (INIS)

    Tessler, G.; Beaudoin, B.R.; Beggs, W.J.; Freeman, L.B.; Schick, W.C. Jr.

    1987-09-01

    A gauge, called the Production Irradiated Fuel Assay Gauge (PIFAG), has been developed and utilized to measure, nondestructively, the fissile fuel content of spent fuel rods from the Light Water Breeder Reactor (LWBR) core. The PIFAG was in operation from November 1983 to May 1987. During this period, assay data were obtained for two irradiated test rods used for initial qualification of the gauge and 524 spent LWBR core rods. A review of PIFAG operations is given, including hot cell operations, calibration, assay operations, and methods used to monitor the data quality and verify the precision and accuracy of the data. The analytical model used to determine the core rod fissile fuel content from the data and the results for the 524 LWBR spent fuel rods are given

  20. Nuclear fuel rod grip with modified diagonal spring structures

    International Nuclear Information System (INIS)

    DeMario, E.E.

    1990-01-01

    This patent describes a spring structure in a nuclear fuel rod grid including a plurality of inner and outer straps being interleaved with one another to form a matrix of hollow cells. Each of the cells is for receiving one fuel rod and being defined by pairs of opposing wall sections of the straps which wall sections are shared with adjacent cells. Each of the cells has a central longitudinal axis, a fuel rod engaging spring structure of resiliently yieldable material being integrally formed on each wall section of the inner straps. The spring structure comprising: a pair of spaced apart opposite outer portions being integrally attached at their outer ends to the respective wall section. The portions extending in alignment with one another and in generally diagonal relation to the direction of the central longitudinal axis of the one cell; and a middle portion disposed between and integrally connected at its outer ends with respective inner ends of the outer portions. The middle portion extending in generally transverse relation to the direction of the central longitudinal axis of the one cell

  1. LOFT fuel rod surface temperature measurement testing

    International Nuclear Information System (INIS)

    Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

    1978-01-01

    Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

  2. Deformation and collapse of zircaloy fuel rod cladding into plenum axial gaps

    International Nuclear Information System (INIS)

    Pfennigwerth, P.L.; Gorscak, D.A.; Selsley, I.A.

    1983-01-01

    To minimize support structure, blanket and reflector fuel rods of the thoria urania-fueled Light Water Breeder Reactor (LWBR) were designed with non-freestanding Zircaloy-4 cladding. An analytical model was developed to predict deformation of unirradiated cladding into axial gaps of fuel rod plenum regions where it is unsupported. This model uses the ACCEPT finite element computer program to calculate elastic-plastic deformation of cladding due to external pressure. The finite element is 20-node, triquadratic, isoparametric, and 3-dimensional. Its curved surface permits accurate modeling of the tube geometry, including geometric nonuniformities such as circumferential wall thickness variation and initial tube out-of-roundness. Progressive increases in axial gap length due to cladding elongation and fuel stack shrinkage are modeled, as are deformations of fuel pellets and stainless steel support sleeves which bound plenum axial gaps in LWBR type blanket fuel rods. Zircaloy-4 primary and secondary thermal creep representations were developed from uniaxial creep testing of fuel rod tubing. Creep response to multi-axial loading is modeled with a variation of Hill's formulation for anisotropic materials. Coefficients accounting for anisotropic thermal creep in Zircaloy-4 tubes were developed from creep testing of externally pressurized tubes having fixed axial gaps in the range 2.5 cm to 5.7 cm and radial clearances over simulated fuel pellets ranging from zero to 0.089 mm. (orig./RW)

  3. Optical coherence tomography for nondestructive evaluation of fuel rod degradation

    International Nuclear Information System (INIS)

    Renshaw, Jeremy B.; Jenkins, Thomas P.; Buckner, Benjamin D.; Friend, Brian

    2015-01-01

    Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such as Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal

  4. Optical coherence tomography for nondestructive evaluation of fuel rod degradation

    Energy Technology Data Exchange (ETDEWEB)

    Renshaw, Jeremy B., E-mail: jrenshaw@epri.com [Electric Power Research Institute, 1300 West WT Harris Blvd., Charlotte, NC 28262 (United States); Jenkins, Thomas P., E-mail: tjenkins@metrolaserinc.com; Buckner, Benjamin D., E-mail: tjenkins@metrolaserinc.com [MetroLaser, Inc., 22941 Mill Creek Drive, Laguna Hills, CA 92653 (United States); Friend, Brian [AREVA, Inc., 3315 Old Forest Road, Lynchburg, VA 24501 (United States)

    2015-03-31

    Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such as Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal.

  5. Pressure drop ana velocity measurements in KMRR fuel rod bundles

    International Nuclear Information System (INIS)

    Yagn, Sun Kyu; Chung, Heung June; Chung, Chang Whan; Chun, Se Young; Song, Chul Wha; Won, Soon Yeun; Chung, Moon Ki

    1990-01-01

    The detailed hydraulic characteristic measurements in subchannels of longitudinally finned rod bundles using one-component LDV(Laser Doppler Velocimeter) were performed. Time mean axial velocity, turbulent intensity, and turbulent micro scales, such as time auto-correlation, Eulerian integral and micro scale, Kolmogorov length and time scale, and Taylor micro length scale were measured. The signals from LDV are inherently more or less discontinuous. The spectra of signals having such intermittent defects can be obtained by the fast Fourier transformation (FFT) of the auto-correlation function. The turbulent crossflow mixing rate between neighboring subchannels and dominant frequencies were evaluated from the measured data. Pressure drop data were obtained for the typical 36-element and 18-element fuel rod bundles fabricated by the design requirement of KMRR fuel and for other type of fuels assembled with 6-fin rods to investigate the fin effects on the pressure drop characteristics

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

    International Nuclear Information System (INIS)

    Nissen, K.L.

    1988-06-01

    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.) [de

  7. Experimental design for HTGR fuel rods

    International Nuclear Information System (INIS)

    Bayne, C.K.

    1975-01-01

    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

  8. Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

    International Nuclear Information System (INIS)

    Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

    1979-01-01

    Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

  9. The Light-Water-Reactor Version of the URANUS Integral fuel-rod code

    Energy Technology Data Exchange (ETDEWEB)

    Labmann, K; Moreno, A

    1977-07-01

    The LWR version of the URANUS code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via URANUS results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analyses of in-pile experiments, they illustrate rather typical and diverse URANUS capabilities. The performance test shows that URANUS is reliable and efficient, thus the code is a most valuable tool in fuel rod analysis work. K. LaBmann developed the LWR version of the URANUS code, material properties were reviewed and supplied by A. Moreno. (Author) 41 refs.

  10. FREC-4A: a computer program to predict fuel rod performance under normal reactor operation

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Izumi, Fumio

    1981-10-01

    The program FREC-4A (Fuel Reliability Evaluation Code-version 4A) is used for predicting fuel rod performance in normal reactor operation. The performance is calculated in accordance with the irradiation history of fuel rods. Emphasis is placed on the prediction of the axial elongation of claddings induced by pellet-cladding mechanical interaction, including the influence of initially preloaded springs inserted in fuel rod lower plenums. In the FREC-4A, an fuel rod is divided into axial segments. In each segment, it is assumed that the temperature, stress and strain are axi-symmetrical, and the axial strain in constant in fuel pellets and in a cladding, though the values in the pellets and in the cladding are different. The calculation of the contact load and the clearance along the length of a fuel rod and the stress and strain in each segment is explained. The method adopted in the FREC-4A is simple, and suitable to predict the deformation of fuel rods over their full length. This report is described on the outline of the program, the method of solving the stiffness equations, the calculation models, the input data such as irradiation history, output distribution, material properties and pores, the printing-out of input data and calculated results. (Kako, I.)

  11. Elliptical cross section fuel rod study II

    International Nuclear Information System (INIS)

    Taboada, H.; Marajofsky, A.

    1996-01-01

    In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab

  12. IFPE/TRIBULATION R1, Fuel Rod Behaviour at High Burnup

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    2002-01-01

    Description: The TRIBULATION (Tests Relative to High Burnup Limitations Arising Normally in LWRs) International Programme started in July 1980 and was organized jointly by BelgoNucleaire and the Nuclear Energy Centre at Mol (CEN/SCK) with the co-sponsorship of 14 participating organizations. The objectives of the programme were twofold. It was primarily a demonstration programme aimed at assessing the fuel rod behaviour at high burn-up, when an earlier transient had occurred in the power plant. The second objective was to investigate the behaviour of different fuel rod designs and manufacturers when subjected to a steady state irradiation history to high burn-up. The first objective was met by irradiating fuel rods under steady state conditions in the BR3 reactor and under transient conditions in BR2. The effect of the transient was determined by comparing data from 4 identical rods tested as follows: i) BR3 irradiation followed by PIE; ii) BR3 irradiation followed by BR2 transient then PIE; iii) BR3 irradiation followed by BR2 transient and re-irradiated in BR3 before PIE; iv) BR3 irradiation and continued BR3 irradiation to maximum burn-up before PIE. The Database contains data from 19 cases using rods fabricated by BelgoNucleaire (BN) (11) and Brown Boveri Reactor GmbH (BBR) (8)

  13. Simulation of fuel rod behaviour during various break LOCAs in PWRs

    International Nuclear Information System (INIS)

    Gadalla, A.A.; El-Fawal, M.M.

    1996-01-01

    During loss of coolant accident (LOCAs) course of events, attention focuses on fuel rod cladding temperature behaviour. In this study, the DRUFAN analytical model and LOBI-MOD2 experimental modeling scheme for fuel rod temperature behaviour during C L-Break LOCA in PWRs, are described and discussed. These models are applied for the investigation of fuel rod cladding temperature behaviour during LOCA blowdown phase. A spectrum of selected values representing small, intermediate and large CL- Break sizes are considered in the predictions. The results of the predictions demonstrated that calculated heater rod temperature at steady state as well as the transient period up to 1000 sec are going in good agreement with the measured values. However above 1000 sec the calculated temperatures are higher than the measured values. This indicates that code predictions in this period are conservative. The results indicated also that, in case of small CL-break LOCA (0.01 A and 0.01 and 0.03 A), the heater rod cladding temperature don't rise above saturation temperature. However, on the top of the heater rod, DNB is occurred in case of 0.03 A CL break, while for 0.01 A break, DNB didn't occur. In case of intermediate and large CL-break; (0.05 A, 0.10 A and 1 A), the results showed that, the heater rod cladding temperature exceeded the saturation temperature and DNB prevailed in upper and intermediate sections of the core. 15 figs., 2 tabs

  14. Experience with a fuel rod enrichment scanner

    International Nuclear Information System (INIS)

    Kubik, R.N.; Pettus, W.G.

    1975-01-01

    This enrichment scanner views all fuel rods produced at B and W's Commercial Nuclear Fuel Plant. The scanner design is derived from the PAPAS System reported by R. A. Forster, H. D. Menlove, and their associates at Los Alamos. The spatial resolution of the system and smoothing of the data are discussed in detail. The cost-effectiveness of multi-detector versus single detector scanners of this general design is also discussed

  15. Heat Transfer Coefficient Variations in Nuclear Fuel Rod Bundles

    International Nuclear Information System (INIS)

    Conner, Michael E.; Holloway, Mary V.

    2007-01-01

    The single-phase heat transfer performance of a PWR nuclear fuel rod bundle is enhanced by the use of mixing vanes attached to the downstream edges of the support grid straps. This improved single-phase performance will delay the onset of nucleate boiling, thereby reducing corrosion and delaying crud-related issues. This paper presents the variation in measured single-phase heat transfer coefficients (HTC) for several grid designs. Then, this variation is compared with observations of actual in-core crud patterns. While crud deposition is a function of a number of parameters including rod heat flux, the HTC is assumed to be a primary factor in explaining why crud deposition is a local phenomenon on nuclear fuel rods. The data from this study will be used to examine this assumption by providing a comparison between HTC variations and crud deposition patterns. (authors)

  16. Development of non-destructive examination system for irradiated fuel rods

    International Nuclear Information System (INIS)

    Sumerling, R.; Goldsmith, L.A.; Cross, M.T.; McKee, F.

    1978-12-01

    The development of non-destructive examination (NDE) system for irradiated fuel rods is described. The system is used for testing rods within a concrete cave and consists of three parts: a fully-automated fuel rod-drive machine, designed for easy maintenance; a series of plug-in NDE modules which fit into the central space provided in the machine, plus optical/TV viewing devices and gamma-scan equipment lined up on the rod; and on electronic control equipment situated outside the concrete shielding. The equipment is at present routinely used for viewing, eddy-current testing, gamma-scanning and diameter measurement of rods. The system is flexible in that additional modules can be added later as they are developed, since there is room for three modules of standard size (about 10cm x 10 cm x 3cm) in the machine or one large module taking the full space. New developments include the use of dual frequency eddy-current testing, which allows much greater discrimination against unwanted signals, and measurement of oxide thickness using a high frequency eddy-current probe. (author)

  17. Results of calculation of WWER-440 fuel rods (Kol`skaya-3 NPP) at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Scheglov, A; Proselkov, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Panin, M; Pitkin, Yu [Kol` skaya NPP, (Russian Federation); Tzibulya, V [AO Mashinostroitelnij Zavod Electrostal (Russian Federation)

    1994-12-31

    Thermal-physical characteristics of fuel rods of two fuel assemblies which were operated within 5 - 8 and 5 - 9 core fuel loadings of the Unit 3 of the Kol`skaya NPP are calculated. They have achieved deep burnup during 4-year (> 46 Mwd/kg U) and 5-year (> 48 Mwd/kg U) fuel cycle. Fuel assemblies have been unloaded off the reactor and subjected to a post-irradiation testing. PIN-mod2 code originally designed for modelling of WWER fuel rod behaviour in a quasi-steady-state operation is used. The average fuel rod in the fuel assembly and the fuel rod with maximum burnup are selected. The preliminary comparison of the calculation results with those of the post-irradiation examination shows a satisfactory agreement. On the basis of the results obtained in the post-irradiation experiments an improvement of the model for calculation of fission gas release and creep of the cladding is planned. The results of the analysis performed indicate that the fuel rod completely preserves its working ability; fuel temperature does not exceed 1300{sup o} C; fission gas release does not exceed 4%; maximum gas pressure inside the cladding at the end of campaign does not exceed 2 MPa. 2 tabs., 11 figs., 5 refs.

  18. Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

    International Nuclear Information System (INIS)

    Murphy, M.F.; Jatuff, F.; Grimm, P.; Seiler, R.; Brogli, R.; Meier, G.; Berger, H.-D.; Chawla, R.

    2006-01-01

    Fuel rods with burnup values beyond 50 GWd/t are characterised by relatively large amounts of fission products and a high abundance of major and minor actinides. Of particular interest is the change in the reactivity of the fuel as a function of burnup and the capability of modern codes to predict this change. In addition, the neutron emission from burnt fuel has important implications for the design of transport and storage facilities. Measurements have been made of the reactivity effects and the neutron emission rates of highly burnt uranium oxide and mixed oxide fuel rod samples coming from a pressurised water reactor (PWR). The reactivity measurements have been made in a PWR lattice in the PROTEUS zero-energy reactor moderated in turn with: water, a water and heavy water mixture and water containing boron. A combined transport flask and sample changer was used to insert the 400 mm long burnt fuel rod segments into the reactor. Both control rod compensation and reactor period methods were used to determine the reactivities of the samples. For the range of burnup values investigated, an interesting exponential relationship has been found between the neutron emission rate and the measured reactivity

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

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza

    2014-01-01

    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)

  20. 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

  1. Reactor core with rod-shaped fuel cells

    International Nuclear Information System (INIS)

    Dworak, A.

    1976-01-01

    The proposal refers to the optimization of the power distribution in a reactor core which is provided with several successive rod-shaped fuel cells. A uniform power output - especially in radial direction - is aimed at. This is achieved by variation of the dwelling periods of the fuel cells, which have, for this purpose, a fuel mixture changing from layer to layer. The fuel cells with the shortest dwelling period are arranged near the coolant inlet side of the reactor core. The dwelling periods of the fuel cells are adapted to the given power distribution. As neighboring cells have equal dwelling periods, the exchange can be performed much easier then with the composition currently known. (UWI) [de

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

    International Nuclear Information System (INIS)

    Roshan Zamir, M.

    2000-01-01

    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 UO 2 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

  3. Stressed and strained state for cermetic-rod-type fuel element

    International Nuclear Information System (INIS)

    Kulikov, I.S.

    1987-01-01

    Calculation technique for designing the stress-strained state of a cermetic rod-type fuel element has been proposed. The technique is based on the time-dependent step-by-step method and the solution of the deformation equilibrium equation for continuous and thick-wall long cylinders at every temporal step by the finite difference method. Additional strains, caused by thermal expansion and radiation swelling, have been taken into account. The transion from the non-contact model to the stiff-contact model has been provided in the case of cladding-fuel gap dissappearing in one or a number of cross-sections along the fuel element height. The method is supplemented by the formula for fuel cans stability estimation in the case of high coolant external pressure. The example of estimation of the cermetic-rod-type fuel elements are considered as an example

  4. Development of a Computer Code for the Estimation of Fuel Rod Failure

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I.H.; Ahn, H.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    Much research has already been performed to obtain the information on the degree of failed fuel rods from the primary coolant activities of operating PWRs in the last few decades. The computer codes that are currently in use for domestic nuclear power plants, such as CADE code and ABB-CE codes developed by Westinghouse and ABB-CE, respectively, still give significant overall errors in estimating the failed fuel rods. In addition, with the CADE code, it is difficult to predict the degree of fuel rod failures during the transient period of nuclear reactor operation, where as the ABB-CE codes are relatively more difficult to use for end-users. In particular, the rapid progresses made recently in the area of the computer hardware and software systems that their computer programs be more versatile and user-friendly. While the MS windows system that is centered on the graphic user interface and multitasking is now in widespread use, the computer codes currently employed at the nuclear power plants, such as CADE and ABB-CE codes, can only be run on the DOS system. Moreover, it is desirable to have a computer code for the fuel rod failure estimation that can directly use the radioactivity data obtained from the on-line monitoring system of the primary coolant activity. The main purpose of this study is, therefore, to develop a Windows computer code that can predict the location, the number of failed fuel rods,and the degree of failures using the radioactivity data obtained from the primary coolant activity for PWRs. Another objective is to combine this computer code with the on-line monitoring system of the primary coolant radioactivity at Kori 3 and 4 operating nuclear power plants and enable their combined use for on-line evaluation of the number and degree of fuel rod failures. (author). 49 refs., 85 figs., 30 tabs.

  5. Technical measurement of small fission gas inventory in fuel rod with laser puncturing system

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Kim, Sung Ryul; Lee, Byoung Oon; Yang, Yong Sik; Baek, Sang Ryul; Song, Ung Sup

    2012-01-01

    The fission gas release cause degradation of fuel rod. It influences fuel temperature and internal pressure due to low thermal conductivity. Therefore, fission gas released to internal void of fuel rod must be measured with burnup. To measure amount of fission gas, fuel rod must be punctured by a steel needle in a closed chamber. Ideal gas law(PV=nRT) is applied to obtain atomic concentration(mole). Steel needle type is good for large amount of fission gas such as commercial spent fuel rod. But, some cases with small fuel rig in research reactor for R/D program are not available to use needle type because of large chamber volume. The laser puncturing technique was developed to solve measurement of small amount of fission gas. This system was very rare equipment in other countries. Fine pressure gage and strong vacuum system were installed, and the chamber volume was reduced at least. Fiber laser was used for easy operation

  6. Sensitivity Analysis of Gap Conductance for Heat Split in an Annular Fuel Rod

    International Nuclear Information System (INIS)

    Chun, Kun Ho; Chun, Tae Hyun; In, Wang Kee; Song, Keun Woo

    2006-01-01

    To increase of the core power density in the current PWR cores, an annular fuel rod was proposed by MIT. This annular fuel rod has two coolant channels and two cladding-pellet gaps unlike the current solid fuel rod. It's important to predict the heat split reasonably because it affects coolant enthalpy rise in each channel and Departure from Nuclear Boiling Ratio (DNBR) in each channel. Conversely, coolant conditions affect fuel temperature and heat split. In particular if the heat rate leans to either inner or outer channel, it is out of a thermal equilibrium. To control a thermal imbalance, placing another gap in the pellet is introduced. The heat flow distribution between internal and external channels as well as fuel and cladding temperature profiles is calculated with and without the fuel gap between the inner and outer pellets

  7. Destructive examination of 3-cycle LWR fuel rods from Turkey Point Unit 3 for the Climax-Spent Fuel Test

    International Nuclear Information System (INIS)

    Atkin, S.D.

    1981-06-01

    The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reactor fuel rods with similar burnups (28 GWd/MTU) and operating histories

  8. A Study on the Structural Integrity Issues of a Dual-Cooled Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Kang-Hee; Lee, Young-Ho; Yoon, Kyung-Ho; Kim, Jae-Yong; Song, Kun-Woo [Korea Atomic Energy Research Institute, 1045 Daedeokdaero Yuseong Daejeon 305-353 (Korea, Republic of)

    2009-06-15

    A dual-cooled fuel rod has an internal coolant flow passage in addition to the external one. A remarkable power up-rate can be achieved due to the increased surface area, which may draw great interests from the fuel researchers, designers and vendors. However, it requires effective resolution to the difficult technical issues when a fuel assembly is to be realized. It becomes much more difficult if a tough boundary condition needs to be satisfied such as a compatibility with the existing reactor internal structures. This kind of challenge is tackled through a national R and D project in Korea: to develop the structural components of a dual-cooled fuel that should be compatible with the current OPR 1000 (Korea Standard Nuclear Power Plant) internal structures. Fuel rod supporting structures, top and bottom end pieces and guide tubes are the components. Besides, the fuel rod components have to be developed as well since the fuel rod's geometry becomes much different from the conventional rod's one. The dimension change may well affect the above mentioned structural components. As a part of the work, structural integrity of the components of a dual-cooled fuel rod is studied in this paper. The investigated topics are: i) the thickness determination of a cladding tube (especially outer tube of a large diameter), ii) vibration issue of an inner cladding tube, iii) design concern of plenum spring and spacer. The cladding thickness issue arises due to the increased outside diameter of a fuel rod, which is caused by an internal flow passage formation. Among the criteria for the thickness determination, an elastic buckling criteria was focused on. Theoretical background for the well-known formula (such as a stability problem) was revisited. Verification tests were carried out independently with using a cladding tube of PHWR fuel rod. Results showed that the formula was not conservative to apply for the cladding thickness determination. Minimum thickness for the

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

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2001-01-01

    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

  10. An investigation on the irradiation behavior of atomized U-Mo/Al dispersion rod fuels

    International Nuclear Information System (INIS)

    Park, J.M.; Ryu, H.J.; Lee, Y.S.; Lee, D.B.; Oh, S.J.; Yoo, B.O.; Jung, Y.H.; Sohn, D.S.; Kim, C.K.

    2005-01-01

    The second irradiation fuel experiment, KOMO-2, for the qualification test of atomized U-Mo dispersion rod fuels with U-loadings of 4-4.5 gU/cc at KAERI was finished after an irradiation up to 70 at% U 235 peak burn-up and subjected to the IMEF (Irradiation material Examination Facility) for a post-irradiation analysis in order to understand the fuel irradiation performance of the U-Mo dispersion fuel. Current results for PIE of KOMO-2 revealed that the U-Mo/Al dispersion fuel rods exhibited a sound performance without any break-away swelling, but most of the fuel rods irradiated at a high linear power showed an extensive formation of the interaction phase between the U-Mo particle and the Al matrix. In this paper, the analysis of the PIE results, which focused on the diffusion related microstructures obtained from the optical and EPMA (Electron Probe Micro Analysis) observations, will be presented in detail. And a thermal modeling will be carried out to calculate the temperature of the fuel rod during an irradiation. (author)

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

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1992-03-01

    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)

  12. Development of program for evaluating the temperature of Zr-U metallic fuel rod

    International Nuclear Information System (INIS)

    Chun, J. S.; Lee, B. H.; Ku, Y. H.; Oh, J. Y.; Im, J. S.; Sohn, D. S.

    2003-01-01

    A code for evaluating the temperature of Zr-U metallic rod has been developed. Finite element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for the Zr-U metallic fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes

  13. Linear variable differential transformer and its uses for in-core fuel rod behavior measurements

    International Nuclear Information System (INIS)

    Wolf, J.R.

    1979-01-01

    The linear variable differential transformer (LVDT) is an electromechanical transducer which produces an ac voltage proportional to the displacement of a movable ferromagnetic core. When the core is connected to the cladding of a nuclear fuel rod, it is capable of producing extremely accurate measurements of fuel rod elongation caused by thermal expansion. The LVDT is used in the Thermal Fuels Behavior Program at the U.S. Idaho National Engineering Laboratory (INEL) for measurements of nuclear fuel rod elongation and as an indication of critical heat flux and the occurrence of departure from nucleate boiling. These types of measurements provide important information about the behavior of nuclear fuel rods under normal and abnormal operating conditions. The objective of the paper is to provide a complete account of recent advances made in LVDT design and experimental data from in-core nuclear reactor tests which use the LVDT

  14. Fuel loading and control rod patterns optimization in a BWR using tabu search

    International Nuclear Information System (INIS)

    Castillo, Alejandro; Ortiz, Juan Jose; Montes, Jose Luis; Perusquia, Raul

    2007-01-01

    This paper presents the QuinalliBT system, a new approach to solve fuel loading and control rod patterns optimization problem in a coupled way. This system involves three different optimization stages; in the first one, a seed fuel loading using the Haling principle is designed. In the second stage, the corresponding control rod pattern for the previous fuel loading is obtained. Finally, in the last stage, a new fuel loading is created, starting from the previous fuel loading and using the corresponding set of optimized control rod patterns. For each stage, a different objective function is considered. In order to obtain the decision parameters used in those functions, the CM-PRESTO 3D steady-state reactor core simulator was used. Second and third stages are repeated until an appropriate fuel loading and its control rod pattern are obtained, or a stop criterion is achieved. In all stages, the tabu search optimization technique was used. The QuinalliBT system was tested and applied to a real BWR operation cycle. It was found that the value for k eff obtained by QuinalliBT was 0.0024 Δk/k greater than that of the reference cycle

  15. Burnable poison rod for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Funk, C.E.; Oneufer, A.S.

    1984-01-01

    A burnable poison rod for use in a nuclear reactor fuel assembly which includes concentrically disposed rods having an annular space therebetween which extends the full length of the rods. The inner rod is hollow to permit circulation of coolant therethrough. Annular burnable poison pellets are positioned in the annular space which is closed at both ends by plugs. A spring clip is located in the plenum space above the pellet stack in the rods. The spring clip is of cylindrical configuration having a gap in the material which provides two ends adapted to be squeezed toward each other. A cross section of the clip shows that its ends contain alternating flat and round edges, the round edges conforming to the outer rod inner surface to provide a retentive force which is releasably applied to the pellet stack as it grows during operation in a reactor

  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. The Study on Radioactive Nuclide Distributions within a Fuel Rod by Tomographic Gamma Scanning Method

    International Nuclear Information System (INIS)

    Quanhu, Zhang; Lee, H. K.; Hong, K. P.; Choo, Y. S.; Kim, D. S.

    2005-06-01

    Based on the specified need of the IMEF, the feasibility of Tomographic Gamma Scanning (TGS) technique has been investigated for its potential for non-destructive gamma scanning measurements of irradiated fuel rods. TGS technique has been developed for determining some radioactive isotopes' distributions of a fuel rod in hot cell. The results obtained from the simulation model extracting from real gamma scanning experimental condition in this work by new developed computer simulation codes confirmed that the gamma emission TGS technique has potential for determination of radioactive isotopes' distributions of a fuel rod. In order to verify the simulation codes, we have designed several computation schemes for both 3 by 3 and 10 by 10 fuel rod model under present situation at M1 hot cell in IMEF. The results which relative errors are less than 10% show that we have simulated and implemented determination of radioactive isotopes' distributions on simulated fuel rod by TGS technique successfully

  18. Characterization of irradiated fuel rods using pulsed eddy current techniques

    International Nuclear Information System (INIS)

    Martin, M.R.; Francis, W.C.

    1975-11-01

    A number of irradiated fuel rods and unfueled zircaloy cladding tubes (''water tubes'') were obtained from the Saxton reactor through arrangements with the Westinghouse Electric Corporation for use in subsequent irradiation effects and fuel behavior programs. A comprehensive nondestructive and corroborative destructive characterization program was undertaken on these fuel rods and tubes by ANC to provide baseline data on their characteristics prior to further testing and for comparison against post-post data. This report deals primarily with one portion of the NDT program performed remotely in the hot cells. The portion of interest in this paper is the pulsed eddy current inspection used in the nondestructive phase of the work. 6 references

  19. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    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)

  20. Computer simulation of the behaviour and performance of a CANDU fuel rod

    International Nuclear Information System (INIS)

    Marino, A.C.

    1997-01-01

    At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

  1. Connection between end plates and rods in a BWR fuel element

    International Nuclear Information System (INIS)

    Cali', G.P.

    1975-01-01

    The problem of the connection between the end plates and the rods of a BWR fuel element is analytically formulated. The behaviour of the springs coupling the rods with the upper plate is analyzed with particular detail since the deformation of these springs affects the forces at the interface of the fuel element structure components. A tool is given to design the springs according to some considerations regarding the mechanical strength of the interacting components as well as the influence of the possible geometrical unevennes of the system that can arise during the fuel element lifetime. (Cali', G.P.)

  2. Development of Fuel ROd Behavior Analysis code (FROBA) and its application to AP1000

    International Nuclear Information System (INIS)

    Yu, Hongxing; Tian, Wenxi; Yang, Zhen; SU, G.H.; Qiu, Suizheng

    2012-01-01

    Highlights: ► A Fuel ROd Behavior Analysis code (FROBA) has been developed. ► The effects irradiation and burnup has been considered in FROBA. ► The comparison with INL’s results shows a good agreement. ► The FROBA code was applied to AP1000. ► Peak fuel temperature, gap width, hoop strain, etc. were obtained. -- Abstract: The reliable prediction of nuclear fuel rod behavior is of great importance for safety evaluation of nuclear reactors. In the present study, a thermo-mechanical coupling code FROBA (Fuel ROd Behavior Analysis) has been independently developed with consideration of irradiation and burnup effects. The thermodynamic, geometrical and mechanical behaviors have been predicted and were compared with the results obtained by Idaho National Laboratory to validate the reliability and accuracy of the FROBA code. The validated code was applied to analyze the fuel behavior of AP1000 at different burnup levels. The thermal results show that the predicted peak fuel temperature experiences three stages in the fuel lifetime. The mechanical results indicate that hoop strain at high power is greater than that at low power, which means that gap closure phenomenon will occur earlier at high power rates. The maximum cladding stress meets the requirement of yield strength limitation in the entire fuel lifetime. All results show that there are enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The FROBA code is expected to be applied to deal with more complicated fuel rod scenarios after some modifications.

  3. Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Majed, M.; Morback, G.; Wiman, P. [ABB Atom AB, Vasteras (Sweden)] [and others

    1995-09-01

    The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give large advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.

  4. Characterization of LWR fuel rod irradiations with power transients in the BR2 reflector

    International Nuclear Information System (INIS)

    Ponsard, B.; Bodart, S.; Meer, K. van der; Raedt, C. de

    1996-01-01

    Fuel rod irradiations in reflector positions of the materials testing reactor BR2 are becoming increasingly important. A typical example is that of irradiation devices containing single LWR fuel rods, to be tested in the framework of a new international fuel investigation and development programme. Some of the irradiations will comprise power transients with central fuel melting (at 2800 deg. C), the power increase being obtained by decreasing the pressure in a He-3 neutron absorbing screen and/or by varying the BR2 reactor operating power. A total power variation by a factor of at least 2.5 in the fuel rod irradiated could thus be achieved. In some of the rods, central temperature measurements (up to 2000 deg. C) will be carried out. Both fresh and pre-irradiated fuel rods are concerned in the programme. For these irradiations, the accurate knowledge of the neutron-induced fission heating and of the gamma heating is required, as one of the purposes of the programme consists in establishing the correlation among the thermal conductivity, the burn-up and the irradiation temperature. Calibration work among various measuring methods and between measurements and one- and two-dimensional calculations is being pursued. (author). 10 refs, 15 figs, 3 tabs

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

    International Nuclear Information System (INIS)

    Marino, A.C.

    2000-01-01

    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)

  6. Probabilistic fuel rod analyses using the TRANSURANUS code

    Energy Technology Data Exchange (ETDEWEB)

    Lassmann, K; O` Carroll, C; Laar, J Van De [CEC Joint Research Centre, Karlsruhe (Germany)

    1997-08-01

    After more than 25 years of fuel rod modelling research, the basic concepts are well established and the limitations of the specific approaches are known. However, the widely used mechanistic approach leads in many cases to discrepancies between theoretical predictions and experimental evidence indicating that models are not exact and that some of the physical processes encountered are of stochastic nature. To better understand uncertainties and their consequences, the mechanistic approach must therefore be augmented by statistical analyses. In the present paper the basic probabilistic methods are briefly discussed. Two such probabilistic approaches are included in the fuel rod performance code TRANSURANUS: the Monte Carlo method and the Numerical Noise Analysis. These two techniques are compared and their capabilities are demonstrated. (author). 12 refs, 4 figs, 2 tabs.

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

    International Nuclear Information System (INIS)

    Joung, Changyoung; Hong, Jintae; Kim, Kahye; Huh, Sungho

    2014-01-01

    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

  8. 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

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

    International Nuclear Information System (INIS)

    Viebrock, J.M.

    1981-09-01

    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

  10. Analysis of neutronic parameters related to reduction in fuel rod diameter for Angra-1 reactor fuel elements

    International Nuclear Information System (INIS)

    Faria, Eduardo F.; Sadde, Luciano M.; Sakai, Massao; Gomes, Sydney da S.

    2000-01-01

    The actual fuel element design for Angra-1 PWR satisfies in a very conservative way the design limits established for the critical heat flux as well as for the energy stored in the fuel rod. However, that is not an optimized design under neutronic considerations. The conservative ratio of the H and U atomic densities gives rise to a harder neutron spectrum which reduces its reactivity. In this report, a reduction in fuel rod diameters has been analyzed, keeping however the same rod pitch for geometrical compatibility reasons. By increasing the H/U ratio it is possible to obtain a net gain in reactivity. The optimized diameter in its turn should not jeopardize the reactor safety requirements. The actual trends of the nuclear industry is to extend the cycles and the enrichment by using advanced fuel design. It must be emphasized that this design change gives rise to economical advantages, for example, reduced costs for uranium utilization and enrichment with a net gain in reactivity. (author)

  11. Process for automatic filling of nuclear fuel rod cans

    International Nuclear Information System (INIS)

    Bezold, H.

    1977-01-01

    A drying section is inserted in the production line for the automation of the filling process for fuel rods with nuclear fuel pellets. The pellets are taken in a drum magazine to a drying furnace and then pushed out one after the other into the can to be filled. (TK) [de

  12. 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.

  13. FEMAXI-III, a computer code for fuel rod performance analysis

    International Nuclear Information System (INIS)

    Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.

    1983-01-01

    This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)

  14. Liquid-metal fast breeder reactor fuel rod performance and modeling at high burnup

    International Nuclear Information System (INIS)

    Verbeek, P.; Toebbe, H.; Hoppe, N.; Steinmetz, B.

    1978-01-01

    The fuel rod modeling codes IAMBUS and COMETHE were used in the analysis and interpretation of postirradiation examination results of mixed-oxide fuel pins. These codes were developed in the framework of the SNR-300 research and development (R and D) program at Interatom and Belgonucleaire, respectively. SNR-300 is a liquid-metal fast breeder reactor demonstration plant designed and presently constructed in consortial cooperation by Germany, Belgium, and the Netherlands. RAPSODIE I, the two-bundle irradiation experiment, was irradiated in the French test FBR RAPSODIE FORTISSIMO and is one of the key irradiation experiments within the SNR-300 R and D program. The comparison of code predictions with postirradiation examination results concentrates on clad diameter expansions, clad total axial elongations, fuel differential and total axial elongations, fuel restructuring, and fission gas release. Fuel rod modeling was considered in the light of benchmarking of the codes, and there was consideration of fuel rod design for operation at low and high burnup

  15. The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

    International Nuclear Information System (INIS)

    Rohde, U.

    1998-01-01

    A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported. Fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D model. (author)

  16. Conceptual design of control rod regulating system for plate type fuels of Triga-2000 reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Saminto

    2016-01-01

    Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor has been made. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor was made with refer to study result of instrument and control system which is used in BATAN'S reactor. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor consist of 4 segments that is control panel, translator, driver and display. Control panel is used for regulating, safety and display control rod, translator is used for signal processing from control panel, driver is used for driving control rod and display is used for display control rod level position. The translator was designed in 2 modes operation i.e operation by using PLC modules and IC TTL modules. These conceptual design can be used as one of reference of control rod regulating system detail design. (author)

  17. 'THERMOST' for analysing thermo-structural behaviour of LWR fuel rods under PCI conditions

    International Nuclear Information System (INIS)

    Nuno, H.; Ogawa, S.; Kobayashi, H.

    1983-01-01

    As a method for evaluating fuel rod performance under power ramping or load following operations, the combined FROST/ THERMOST system has been developed and brought into practical use. FROST was presented at the IAEA Blackpool Meeting in 1978, and THERMOST is the subject of this paper. The major purpose of THERMOST is to analyse very detailed thermal and structural fuel behaviour in a rather localised part of the fuel rod whereas FROST deals with whole rod general performance. The code handles two-dimensional thermal and structural analyses simultaneously by using a finite element method, in axial section or in lateral section. It consists of a fundamental FEM system of generalised constitution, and a surrounding subroutine system which characterises fuel behaviour, such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer element (six kinds), and structural analysis by axisymmetric ring and lateral plane element (six kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping conditions is presented with some in-pile test data. (author)

  18. 29 CFR 1910.253 - Oxygen-fuel gas welding and cutting.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 5 2010-07-01 2010-07-01 false Oxygen-fuel gas welding and cutting. 1910.253 Section 1910..., DEPARTMENT OF LABOR OCCUPATIONAL SAFETY AND HEALTH STANDARDS Welding, Cutting and Brazing § 1910.253 Oxygen-fuel gas welding and cutting. (a) General requirements—(1) Flammable mixture. Mixtures of fuel gases...

  19. Acoustic sensor for in-pile fuel rod fission gas release measurement

    International Nuclear Information System (INIS)

    Fourmentel, D.; Villard, J. F.; Ferrandis, J. Y.; Augereau, F.; Rosenkrantz, E.; Dierckx, M.

    2009-01-01

    We have developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: - fuel temperature with a centre-line thermocouple type C, - internal pressure with a specific counter-pressure sensor, - fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced online from this measurement. The principle of this acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in materials testing reactors conditions. For this purpose some technical points have been studied in details: - fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, - size of the gas cavity to avoid any perturbation of the acoustic waves, - miniaturization of the sensor because of narrow in-pile experimental devices

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

    International Nuclear Information System (INIS)

    Leech, W.J.; Kaiser, R.S.

    1980-01-01

    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. Results from In-pile experiments on LWR fuel rod behavior under LOCA conditions with unirradiated rods

    International Nuclear Information System (INIS)

    Sepold, L.; Karb, E.H.; Pruessmann, M.

    1981-06-01

    This report summarizes the results of the FR2-in-pile tests at KfK (Kernforschungszentrum Karlsruhe) with unirradiated test rods. The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated single rods of a PWR design in the DK loop of the FR2 reactor. The main parameter of the test program was the burnup, ranging from 2.500 to 35.000 MWd/t. The program with unirradiated specimens comprised the series A and B with a total of 14 tests. (orig.) [de

  2. PIN99W, Modelling of VVER and PWR Fuel Rod Thermomechanical Behaviour

    International Nuclear Information System (INIS)

    Valach, M.; Strizhov, P.; Svoboda, R.

    2000-01-01

    1 - Description of program or function: The Code is developed to describe fuel rod thermomechanical behaviour in operational conditions. The main goal of this code is to calculate fuel temperature, gap conductivity, fission gas release and inner gas pressure. 2 - Methods: - fuel rod temperature response is solved by using one-dimensional finite element method combined with weighted residuals method; - the code involves models describing physical phenomena typical for the fuel irradiated in Light Water Power Reactors (densification, restructuring, fission gas release, swelling and relocation) ; - this code is updated and improves PIN-micro code. 3 - Restrictions on the complexity of the problem: - simplified mechanistic solution; - only steady-state solution; - no cladding failure criterion; - no model for axial fuel-cladding interaction

  3. Estimation of the core-wide fuel rod damage during a LWR LOCA

    International Nuclear Information System (INIS)

    Mattila, L.; Sairanen, R.; Stengaard, J.-O.

    1975-01-01

    The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

  4. The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

    International Nuclear Information System (INIS)

    Rohde, U.

    2001-01-01

    A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises a detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. A creeping law for time-dependent estimation of plastic deformations is implemented. Metal-water reaction of the cladding material in the high temperature region is considered. The cladding-coolant heat transfer regime map covers the region from one-phase liquid convection to dispersed flow with superheated steam. Special emphasis is put on taking into account the impact of thermodynamic non-equilibrium conditions on heat transfer. For the validation of the model, experiments on fuel rod behaviour during RIAs carried out in Russian and Japanese pulsed research reactors with shortened probes of fresh fuel rods are calculated. Comparisons between calculated and measured results are shown and discussed. It is shown, that the fuel rod behaviour is significantly influenced by plastic deformation of the cladding, post crisis heat transfer with sub-cooled liquid conditions and heat release from the metal-water reaction. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported on. It is demonstrated, that the fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D

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

    International Nuclear Information System (INIS)

    Schienbein, Marcel; Zeh, Peter; Hurtado, Antonio; Rosskamp, Matthias; Mailand, Irene; Bolz, Michael

    2012-09-01

    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

  6. Determination of the axial 235U distribution in target fuel rods

    International Nuclear Information System (INIS)

    Huettig, G.; Bernhard, G.; Niese, U.

    1989-01-01

    The homogenity of the axial 235 U distribution in target fuel rods is an important quality criterion for the production of 99 Mo. The 235 U distribution has been analyzed automatically and nondestructively by measuring the 235 U gamma ray peak at 285.7 keV. For the quantitative assessment a calibration curve was prepared by the help of X-ray fluorescence analysis, colorimetry, and photometric titration. The accuracy of the method is ≤ 1.5% uranium per centimeter of the fuel rod

  7. Interim transfer canister for consolidating nuclear fuel rods

    International Nuclear Information System (INIS)

    Formanek, F.J.

    1987-01-01

    This patent describes a canister for receiving and consolidating a group of uniformly spaced apart nuclear fuel rods, comprising: a rectangular, vertically oriented straight back panel; a pair of oppositely disposed side panels connected perpendicularly to the back panel, having a vertical straight upper portion and an inwardly tapered lower portion; a front panel opposite the back panel and connected to the side panels, having a straight vertical upper portion and inwardly tapered lower portion; whereby the back, side and front panels define a rectangular upper opening at the upper end of the canister and a generally rectangular lower opening at the other end, the lower opening having a cross-sectional area less than one-half that of the upper opening; parallel plate members spanning the canister from the front panel to the back panel, each plate spaced from the other the same uniform distance, the plates extending downwardly into the tapered portion of the canister while remaining spaced above the tapered sidewalls; first base means at the lower end of the canister, removably mounted and having an oblique orientation generally downward from the front panel to the back panel, for guiding the fuel rods to be inserted preferentially toward the lower portion of the back panel; and second base means removably mounted within the canister below first base means and oriented transversely to the longitudinal extent of the canister, for supporting the fuel rods when the first base means is removed from the canister

  8. Large eddy simulation of a fuel rod subchannel

    International Nuclear Information System (INIS)

    Mayer, Gusztav

    2007-01-01

    In a VVER-440 reactor the measured outlet temperature is related to fuel limit parameters and the power upgrading plans of VVER-440 reactors motivated us to obtain more information on the mixing process of the fuel assemblies. In a VVER-440 rod bundle the fuel rods are arranged in triangular array. Measurement shows (Krauss and Meyer, 1998) that the classical engineering approach, which tries to trace the characterization of such systems back to equivalent (hydraulic diameter) pipe flows, does not give reasonable results. Due to the different turbulence characteristics, the mixing is more intensive in rod bundles than it would be expected based on equivalent pipe flow correlations. As a possible explanation of the high mixing, secondary flow was deduced from measurements by several experimentalists (Trupp and Azad, 1975). Another candidate to explain the high mixing is the so-called flow pulsation phenomenon (Krauss and Meyer, 1998). In this paper we present subchannel simulations (Mayer et al. 2007) using large eddy simulation (LES) methodology and the lattice Boltzmann method (LBM) without the spacers at Reynolds number 21000. The simulation results are compared with the measurements of Trupp and Azad (1975). The mean axial velocity profile shows good agreement with the measurement data. Secondary flow has been observed directly in the simulation results. Reasonable agreement has been achieved for most Reynolds stresses. Nevertheless, the calculated normal stresses show small, but systematic deviation from the measurement data. (author)

  9. Technique of manufacturing specimen of irradiated fuel rods

    International Nuclear Information System (INIS)

    Min, Duck Seok; Seo, Hang Seok; Min, Duck Kee; Koo, Dae Seo; Lee, Eun Pyo; Yang, Song Yeol

    1999-04-01

    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

  10. PWR fuel rod corrosion in Japan

    International Nuclear Information System (INIS)

    Inoue, S.; Mori, K.; Murata, K.; Kobasyashi, S.

    1997-01-01

    Many particular appearance were observed on the fuel rod surfaces during fuel inspection at reactor outage in 1991. The appearances looked like small black circular nodules. The size was approximately 1 mm. This kind of appearances were found on fuel rods of which burnup exceeded approximately 30 GWd/t and at the second or third spans of the fuel assembly from the top. In order to clarify the cause, PIE was performed. The black nodules were confirmed to be oxide film spalling by visual inspection. Maximum oxide film thickness was 70 μm and spalling was observed where oxide thickness exceeded 40 t0 50 μm. Oxide film thickness was greater than expected. Many small pores were found in the oxide film when the oxide film had become thicker. Many circumferential cracks were also found in the film. It was speculated that these cracks caused the spalling of the oxide film. Hydride precipitates were mainly oriented circumferentially. Dense hydrides were observed near the outer rim of the cladding. No concentrated hydrides were observed near the spalling area. Maximum hydrogen content was 315 ppm. It was confirmed that the results of tensile test showed no significant effects by corrosion. The mechanism of accelerated corrosion was studied in detail. Water chemistry during irradiation was examined. Lithium content was maintained below 2.2 ppm. pH value was kept between 6.9 and 7.2. There was no anomalies in water chemistry during reactor operation. Cladding fabrication record clarified that heat treatment parameter was smaller than the optimum value. In Japan, heat treatment of the cladding was already optimized by improved fabrication process. Also chemical composition optimization of the cladding, such as low Tin and high Silicon content, was adopted for high burnup fuel. These remedies has already reduced fuel cladding corrosion and we believe we have solved this problem. (author). 6 figs, 1 tab

  11. PWR fuel rod corrosion in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, S [Kansai Electric Power Co., Inc., Osaka (Japan); Mori, K; Murata, K; Kobasyashi, S [Nuclear Fuel Industries, Ltd, Osaka (Japan)

    1997-02-01

    Many particular appearance were observed on the fuel rod surfaces during fuel inspection at reactor outage in 1991. The appearances looked like small black circular nodules. The size was approximately 1 mm. This kind of appearances were found on fuel rods of which burnup exceeded approximately 30 GWd/t and at the second or third spans of the fuel assembly from the top. In order to clarify the cause, PIE was performed. The black nodules were confirmed to be oxide film spalling by visual inspection. Maximum oxide film thickness was 70 {mu}m and spalling was observed where oxide thickness exceeded 40 t0 50 {mu}m. Oxide film thickness was greater than expected. Many small pores were found in the oxide film when the oxide film had become thicker. Many circumferential cracks were also found in the film. It was speculated that these cracks caused the spalling of the oxide film. Hydride precipitates were mainly oriented circumferentially. Dense hydrides were observed near the outer rim of the cladding. No concentrated hydrides were observed near the spalling area. Maximum hydrogen content was 315 ppm. It was confirmed that the results of tensile test showed no significant effects by corrosion. The mechanism of accelerated corrosion was studied in detail. Water chemistry during irradiation was examined. Lithium content was maintained below 2.2 ppm. pH value was kept between 6.9 and 7.2. There was no anomalies in water chemistry during reactor operation. Cladding fabrication record clarified that heat treatment parameter was smaller than the optimum value. In Japan, heat treatment of the cladding was already optimized by improved fabrication process. Also chemical composition optimization of the cladding, such as low Tin and high Silicon content, was adopted for high burnup fuel. These remedies has already reduced fuel cladding corrosion and we believe we have solved this problem. (author). 6 figs, 1 tab.

  12. Core design and fuel rod analyses of a super fast reactor with high power density

    International Nuclear Information System (INIS)

    Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. 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. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. 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 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, 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 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

  13. Analytical model for calculation of the thermo hydraulic parameters in a fuel rod assembly

    Energy Technology Data Exchange (ETDEWEB)

    Cesna, B., E-mail: benas@mail.lei.l [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos g. 3, LT-44403 Kaunas (Lithuania)

    2010-11-15

    Research highlights: {yields} Proposed calculation model can be used for rapid calculation of the bundles with rods spaced by wire wrapping or honey type spacer grids. {yields} Model estimate three flow cross mixture mechanisms. {yields} Program DARS is enable to analyses experimental results. - Abstract: The paper presents the procedure of the cellular calculation of thermo hydraulic parameters of a single-phase gas flow in a fuel rod assembly. The procedure is implemented in the DARS program. The program is intended for calculation of the distribution of the gaseous coolant parameters and wall temperatures in case of arbitrary, geometrically specified, arrangement of the rods in fuel assembly and in case of arbitrary, functionally specified in space, heat release in the rods. In mathematical model the flow cross-section of the channel of intricate shape is conventionally divided to elementary cells formed by straight lines, which connect the centers of rods. Within the limits of a single cell the coolant parameters and the temperature of the corresponding part of the rod surface are assumed constant. The entire fuel assembly is viewed as a system of parallel interconnected channels. Program DARS is illustrated by calculation of a temperature mode of 85-rod assembly with spacers of wire wrapping on the rods.

  14. Recent improvements in modelling fission gas release and rod deformation on metallic fuel in LMR

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, Byoung-Oon; Kim, Young Jin

    2000-01-01

    Metallic fuel design is a key feature to assure LMR core safety goals. To date, a large effort has been devoted to the development of the MACSIS code for metallic fuel rod design and the evaluation of operational limits under irradiation conditions. The updated models of fission gas release, fuel core swelling, and rod deformation are incorporated into the correspondence routines in MACSIS MOD1. The MACSIS MOD1 which is a new version of MACSIS, has been partly benchmarked on FGR, fuel swelling and rod deformation comparing with the results of U-Zr and U-Pu-Zr metal fuels irradiated in LMRs. The MACSIS MOD1 predicts, relatively well, the absolute magnitudes and trends of the gas release and rod deformations depending on burn-up, and it gives better agreement with the experimental data than the previous predictions of MACSIS and the results of the empirical model

  15. HCPWR type fuel elements design with respect to its control rods

    International Nuclear Information System (INIS)

    Abbate, M.J.; Sbaffoni, M.M.; Patino, N.E.; Torasso, O.

    1992-01-01

    The high conversion reactors (HCPWR) can improve the nuclear fuel utilization. One of its present problems is the optimization of the control rods' worth and its relationship with the void coefficient. This investigation means, starting from one reference's design, to analyze several possibilities on the number and distribution of the control rods. As main result, one design including 24 control rods in an optimized distribution, is recommended. (author)

  16. A probabilistic design method for LMFBR fuel rods

    International Nuclear Information System (INIS)

    Peck, S.O.; Lovejoy, W.S.

    1977-01-01

    Fuel rod performance analyses for design purposes are dependent upon material properties, dimensions, and loads that are statistical in nature. Conventional design practice accounts for the uncertainties in relevant parameters by designing to a 'safety factor', set so as to assure safe operation. Arbitrary assignment of these safety factors, based upon a number of 'worst case' assumptions, may result in costly over-design. Probabilistic design methods provide a systematic way to reflect the uncertainties in design parameters. PECS-III is a computer code which employs Monte Carlo techniques to generate the probability density and distribution functions for time-to-failure and cumulative damage for sealed plenum LMFBR fuel rods on a single rod or whole core basis. In Monte Carlo analyses, a deterministic model (that maps single-valued inputs into single-valued outputs) is coupled to a statistical 'driver'. Uncertainties in the input are reflected by assigning probability densities to the input parameters. Dependent input variables are considered multivariate normal. Independent input variables may be arbitrarily distributed. Sample values are drawn from these input densities, and a complete analysis is done by the deterministic model to generate a sample point in the output distribution. This process is repeated many times, and the number of times each output value occurs is accumulated. The probability that some measure of rod performance will fall within given limits is estimated by the relative frequency with which the Monte Carlo samples fall within tho

  17. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author)

  18. Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

    International Nuclear Information System (INIS)

    Casal, V.; Malang, S.; Rust, K.

    1982-10-01

    For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

  19. Results of VVER fuel rods tests in the MIR.M1 reactor under power cycling conditions

    International Nuclear Information System (INIS)

    Burukin, A.; Izhutov, A.; Ovchinnikov, V.; Kalygin, V.; Markov, D.; Pimenov, Y.; Novikov, V.; Medvedev, A.; Nesterov, B.

    2011-01-01

    The paper presents the main results of the 50 ... 60 MWd/kgU burnup VVER fuel rods tests performed in the MIR.M1 reactor loop facilities under power cycling. The non-destructive PIE results are presented as well. A series of experiments was performed, including overall measurement of fuel rod parameters test, in one of which 300 cycles were done. Irradiation under power cycling conditions and PIE of high-burnup VVER fuel rods showed the following: 1) all fuel rods claddings preserved their integrity under irradiation at linear heat rate (LHR) higher than the NPP operating one; 2) experimental data were obtained on the axial and radial cladding strain and fission gas release (FGR) from 50 ... 60 MWd/kgU burnup VVER-440 and VVER-1000 fuel rods as well as on the kinetics of the change in these parameters and fuel temperature under the power cycling; 3) non-destructive PIE results are in a satisfactory correlation with the data obtained by means of in-pile measurement gages during irradiation. (authors)

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

    International Nuclear Information System (INIS)

    Lopez, J.; Montes, M.; Serrano, J.; Haefner, H.E.

    1984-01-01

    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 m 3 /h) and coolant temperature (max. 60degC). The experimental values are compared to the Tecap-Code results. (author)

  1. Hydrodynamics around a spacer of a VVER-440 fuel rod bundle

    International Nuclear Information System (INIS)

    Mayer, G.; Hazi, G.; Kavran, P.

    2004-01-01

    Recently, an intensive research has been started in our institute, focusing on the hydrodynamics of fuel rod bundles. Numerical computations have been planed to be carried out in a three level bottom-up hierarchy, using direct numerical simulation, large eddy simulation and Reynolds averaged Navier-Stokes approach. Here, we give a description of the numerical method applied for direct numerical and large eddy simulation. We present some preliminary results obtained by the simulation of the flow around a spacer of a VVER-440 fuel rod bundle. (author)

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

    International Nuclear Information System (INIS)

    Thakre, Sachin; Lakshmanan, S.P.; Kulkarni, Vinayak; Pandey, Manmohan

    2009-01-01

    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)

  3. Heat split imbalance study for annular fuel rod

    International Nuclear Information System (INIS)

    He Xiaojun; Ji Songtao; Zhang Yingchao

    2014-01-01

    Annular fuel rod has two gaps at inner and outer side. Under irradiation condition, the dimensional change of pellets is always larger than claddings' due to thermal expansion, swelling and densification, and this tends to enlarge the inner gap and reduce the outer gap. The gap size asymmetry must induce heat split imbalance problem that the heat flux will be larger at outer side of the rod. In this work, computer code AFPAC l.0 is used to simulate this heat split imbalance phenomena. The effect of initial gap size, rod inner pressure, roughness of pellets and cladding is studied, the results reveal that: l) Adjusting initial size of both gaps, reducing inner gap and enlarging outer gap could effectively alleviate heat split imbalance problem; 2) Adjusting the initial roughness of pellets and cladding is another effective approach to reducing heat split imbalance; 3) It seems that changing the rod inner pressure has a little effect on solving the heat flux asymmetry problem. (authors)

  4. Development of compaction technique for spent fuel skeletons

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Kim, Young Hwan; Jung, Jae Hoo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    To increase the utilization of uranium resources contained in the spent fuel, the spent fuel is reused. For this, the spent fuel is dismantled or spent fuel rod is extracted from the spent fuel assembly. When the rod is extracted, the remaining components of spent fuel assembly, so called a NFBC(Non-Fuel Bearing Components), should be compacted for the final disposal. To this end, several companies developed the NFBC compactors. German company, named as GNS has developed the direct compression devices of the NFBCs for the rod consolidation and installed it at the PKA(2) of pilot conditioning plant. B and W (Babcock and Wilcox) in USA adopted cutting method rather than the compression method and developed the special cutting devices of NFBC which can be applied underwater environment. In this study the characteristics of these two methods was investigated, in terms of fabrication cost of devices, maintainability in a high radioactive environment, required power and work volume for operation. Also, the optimal power source is selected by comparing the maximum power versus the work volume for operation. In addition to these, the reduction ratio of the bulk volume is obtained while varying the cutting length of the NFBC through a series of experiments. Based on the results of analysis and experiments, the cutting method after compression is selected as an optimal volume reduction method and its design specification is obtained. 8 refs., 62 figs., 32 tabs. (Author)

  5. 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.

  6. Study and simulation of the rim effect in rep fuel rods

    International Nuclear Information System (INIS)

    Hermitte, B.

    1996-01-01

    The RIM effect has been discovered fifteen years ago during the examination of first irradiated rods at more than 45 gWJ/TU in experimental reactors. The rods observation revealed a continuously degradation of the granular structure in the pellet skin, jointly to the porosity increase in this area. This study proposes a RIM formation and development mechanism for high combustion level. The first part presents the simulation of the fission gases in the fuel fraction concerned by the RIM. In the proposed model the gas bubbles increase is bound to the volume fraction of restructured fuel. This model allows the determination of the pores volume fraction in the fuel, the average size of these pores and the volume distribution of the fission gases between the bubbles and the fuel matrix. (A.L.B.)

  7. Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

    1993-02-01

    This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). 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 energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

  8. Neutronic evaluation of annular fuel rods to assemblies 13 x 13, 14 x 14 and 15 x 15

    International Nuclear Information System (INIS)

    Silva, Raphael H.M.; Ramos, Mario C.; Velasquez, Carlos E.; Silva, Clarysson A.M. da; Pereira, Cláubia; Costa, Antonella L.

    2017-01-01

    Research and development in nuclear reactor field has been proposed a new concept of fuel rod such as annular shape. The design of the annular fuel rods allows the coolant flow through the inner and outer side of it. Such project was proposed as an alternative to the traditional fuel rods used in LWR reactors. This new geometry allows an increase in power density in the reactor core with greater heat transfer from the fuel to the coolant which reduces the temperature in central region of the rod, in which a better configuration and dimension of fuel elements are aimed due to improvement of cooling in possible replacement of PWR traditional rods for annular rods. The aim of this work is to evaluate the neutronic parameters of fuel element with annular fuel rods where three configurations were studied: 13 x 13, 14 x 14 and 15 x 15. The goal is compare the neutronic between the advanced and the standard fuel assembly 16 x 16. In these studies, the external dimension and the moderator to fuel volume ratio (V M /V F ) of standard 16 x 16 is the same in all annular fuels assemblies. The MCNPX 2.6.0 (Monte Carlo N-Particle eXtended – version 2.6.0) code was used in all simulations. After all procedures, the annular fuel assemblies 13 have obtained greater neutronics parameters and were selected to more neutronics simulations. (author)

  9. Neutronic evaluation of annular fuel rods to assemblies 13 x 13, 14 x 14 and 15 x 15

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Raphael H.M.; Ramos, Mario C.; Velasquez, Carlos E.; Silva, Clarysson A.M. da; Pereira, Cláubia; Costa, Antonella L., E-mail: rapha.galo@hotmail.com, E-mail: marc5663@gmail.com, E-mail: carlosvelcab@hotmail.com, E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Research and development in nuclear reactor field has been proposed a new concept of fuel rod such as annular shape. The design of the annular fuel rods allows the coolant flow through the inner and outer side of it. Such project was proposed as an alternative to the traditional fuel rods used in LWR reactors. This new geometry allows an increase in power density in the reactor core with greater heat transfer from the fuel to the coolant which reduces the temperature in central region of the rod, in which a better configuration and dimension of fuel elements are aimed due to improvement of cooling in possible replacement of PWR traditional rods for annular rods. The aim of this work is to evaluate the neutronic parameters of fuel element with annular fuel rods where three configurations were studied: 13 x 13, 14 x 14 and 15 x 15. The goal is compare the neutronic between the advanced and the standard fuel assembly 16 x 16. In these studies, the external dimension and the moderator to fuel volume ratio (V{sub M}/V{sub F}) of standard 16 x 16 is the same in all annular fuels assemblies. The MCNPX 2.6.0 (Monte Carlo N-Particle eXtended – version 2.6.0) code was used in all simulations. After all procedures, the annular fuel assemblies 13 have obtained greater neutronics parameters and were selected to more neutronics simulations. (author)

  10. POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

    Directory of Open Access Journals (Sweden)

    H.J. RYU

    2013-12-01

    Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

  11. 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.

  12. Simulation of the fuel rod thermal hydraulic performance during the blow down phase in a PWR

    International Nuclear Information System (INIS)

    Gadelha, J.A.M.

    1982-10-01

    A digital computer code to predict the fuel rod thermalhydraulic performance during a postulated loss-of-coolant accident (LOCA) in the primary circuit of a PWR nuclear power plant is developed. The fuel rod corresponds to that in an average channel in the core. Only the blowdown phase is considered during the accident. The conservation equations of mass, momentum, and energy, and the heat conduction equation are solved to determine the fuel rod conditions during the accident. Finite differences are applied as a numerical method in the solution of the equations modelling the rod and coolant conditions. (Author) [pt

  13. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    OpenAIRE

    ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; th...

  14. Nuclear fuel rod grid spring and dimple structures having chamfered edges for reduced pressure drop

    International Nuclear Information System (INIS)

    De Mario, E.E.

    1990-01-01

    This patent describes a nuclear fuel rod grid including inner and outer straps being interleaved with one another to form a matrix of hollow cells, each cell for receiving one fuel rod and being defined by pairs of opposing wall sections of the straps which wall sections are shared with adjacent cells, each cell having a central longitudinal axis defining a coolant flow direction through the cell, at least fuel rod engaging dimple structure of resiliently yieldable material being integrally formed on each wall section of the inner straps

  15. Apparatus and method for applying an end plug to a fuel rod tube end

    International Nuclear Information System (INIS)

    Rieben, S.L.; Wylie, M.E.

    1987-01-01

    An apparatus is described for applying an end plug to a hollow end of a nuclear fuel rod tube, comprising: support means mounted for reciprocal movement between remote and adjacent positions relative to a nuclear fuel rod tube end to which an end plug is to be applied; guide means supported on the support means for movement; and drive means coupled to the support means and being actuatable for movement between retracted and extended positions for reciprocally moving the support means between its respective remote and adjacent positions. A method for applying an end plug to a hollow end of a nuclear fuel rod tube is also described

  16. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Markgraf, J; Perry, D; Oudaert, J [Commission of the European Communities, Joint Reserach Centre, Petten Establishment, Petten (Netherlands)

    1983-06-01

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  17. TRANSURANUS: A fuel rod analysis code ready for use

    Energy Technology Data Exchange (ETDEWEB)

    Lassmann, K; O` Carroll, C; Van de Laar, J [Commission of the European Communities, Karlsruhe (Germany). European Inst. for Transuranium Elements; Ott, C [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-12-31

    The basic concepts of fuel rod performance codes are discussed. The TRANSURANUS code developed at the Institute for Transuranium Elements, Karlsruhe (GE) is presented. It is a quasi two-dimensional (1{sub 1/2}-D) code designed for treatment of a whole fuel rod for any type of reactor and any situation. The fuel rods found in the majority of test- or power reactors can be analyzed for very different situations (normal, off-normal and accidental). The time scale of the problems to be treated may range from milliseconds to years. The TRANSURANUS code consists of a clearly defined mechanical/mathematical framework into which physical models can easily be incorporated. This framework has been extensively tested and the programming very clearly reflects this structure. The code is well structured and easy to understand. It has a comprehensive material data bank for different fuels, claddings, coolants and their properties. The code can be employed in a deterministic and a statistical version. It is written in standard FORTRAN 77. The code system includes: 2 preprocessor programs (MAKROH and AXORDER) for setting up new data cases; the post-processor URPLOT for plotting all important quantities as a function of the radius, the axial coordinate or the time; the post-processor URSTART evaluating statistical analyses. The TRANSURANUS code exhibits short running times. A new WINDOWS-based interactive interface is under development. The code is now in use in various European institutions and is available to all interested parties. 7 figs., 15 refs.

  18. 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)

  19. Fuel rod failure during film boiling (PCM-1 test in the PBF)

    International Nuclear Information System (INIS)

    Domenico, W.F.; Stanley, C.J.; Mehner, A.S.

    1978-01-01

    The Power-Cooling-Mismatch (PCM) Test, PCM-1 was conducted in the Power Burst Facility (PFB) in March of 1978. The PCM Test Series is being conducted at the Idaho National Engineering Laboratory by EG and G Idaho, Inc., under contract to the USNRC and is designed to characterize the behavior of nuclear fuel rods operating under conditions of high power or low coolant flow or both leading to departure from nucleate boiling. The PCM-1 test was performed to provide in-pile data for a ''worst case'' PCM incident. The objective of this experiment was to study the behavior of a single pressurized water reactor (PWR) fuel rod subjected to a high-power and low flow environment which would result in cladding failure at full power. The ''worst case'' conditions established for the experiment consisted of a rod peak power of 78.7 kW/m and a coolant mass flux of 1356 kg/s.m 2 . Fuel temperatures at the stipulated operating conditions were such that a significant volume of molten fuel was present when failure occurred which produced a high probability of molten fuel-coolant interaction (MFCI) with the possibility of a vapor explosion

  20. Summary of the fuel rod support system (grids) design for LWBR (LWBR development program)

    International Nuclear Information System (INIS)

    Richardson, K.D.

    1979-02-01

    Design features of the fuel rod support system (grids) for the Light Water Breeder Reactor (LWBR) installed in the Shippingport Atomic Power Station, Shippingport, Pennsylvania, are described. The grids are fabricated from AM-350 stainless steel and provide lateral support of the fuel rods in the three regions (seed, blanket, and reflector) of the reactor. A comparison is made of the LWBR grids, whose cells are arranged in triangular-pitched arrays, with rod support systems employed in commercial light water reactors

  1. Nuclear Fuel elements

    International Nuclear Information System (INIS)

    Hirakawa, Hiromasa.

    1979-01-01

    Purpose: To reduce the stress gradient resulted in the fuel can in fuel rods adapted to control the axial power distribution by the combination of fuel pellets having different linear power densities. Constitution: In a fuel rod comprising a first fuel pellet of a relatively low linear power density and a second fuel pellet of a relatively high linear power density, the second fuel pellet is cut at its both end faces by an amount corresponding to the heat expansion of the pellet due to the difference in the linear power density to the adjacent first fuel pellet. Thus, the second fuel pellet takes a smaller space than the first fuel pellet in the fuel can. This can reduce the stress produced in the portion of the fuel can corresponding to the boundary between the adjacent fuel pellets. (Kawakami, Y.)

  2. A parametric thermohydraulic study an advanced pressurized light water reactor with a tight fuel rod lattice

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Hame, W.

    1982-12-01

    A parametric thermohydraulic study for an Advanced Pressurized Light Water Reactor (APWR) with a tight fuel rod lattice has been performed. The APWR improves the uranium utilisation. The APWR core should be placed in a modern German PWR plant. Within this study about 200 different reactors have been calculated. The tightening of the fuel rod lattice implies a decrease of the net electrical output of the plant, which is greater for the heterogeneous reactor than for the homogeneous reactor. APWR cores mean higher core pressure drops and higher water velocities in the core region. The cores tend to be shorter and the number of fuel rods to be higher than for the PWR. At the higher fuel rod pitch to diameter ratios (p/d) the DNB limitation is more stringent than the limitation on the fuel rod linear rating given by the necessity of reflooding after a reactor accident. The contrary is true for the lower p/d ratios. Subcooled boiling in the highest rated coolant channels occurs for the most of the calculated reactors. (orig.) [de

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

    International Nuclear Information System (INIS)

    Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

    1984-01-01

    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

  4. Development of joining techniques for fabrication of fuel rod simulators

    International Nuclear Information System (INIS)

    Moorhead, A.J.; McCulloch, R.W.; Reed, R.W.; Woodhouse, J.J.

    1980-10-01

    Much of the safety-related thermal-hydraulic tests on nuclear reactors are conducted not in the reactor itself, but in mockup segments of a core that uses resistance-heated fuel rod simulators (FRS) in place of the radioactive fuel rods. Laser welding and furnace brazing techniques are described for joining subassemblies for FRS that have survived up to 1000 h steady-state operation at 700 to 1100 0 C cladding temperatures and over 5000 thermal transients, ranging from 10 to 100 0 C/s. A pulsed-laser welding procedure that includes use of small-diameter filler wire is used to join one end of a resistance heating element of Pt-8 W, Fe-22 Cr-5.5 Al-0.5 Co, or 80 Ni-20 Cr (wt %) to a tubular conductor of an appropriate intermediate material. The other end of the heating element is laser welded to an end plug, which in turn is welded to a central conductor rod

  5. Method for placing fuel rods in individual cells, and device for performing this procedure

    Energy Technology Data Exchange (ETDEWEB)

    Jabsen, F S

    1972-10-16

    A lattice grid is described in which an egg box type assembly is formed by metal plates deformed to form spring loaded spacers which retain the fuel pins in their correct position. In order to be able to insert the fuel pins without causing scratches on their surfaces, which could lead to corrosion, the springs are displaced outwards by inserting and rotating a square-sectioned rod with rounded corners which when rotated acts as a cam, pressing the springs out. The springs are held in this position by inserting keys horizontally between the lattice plates, through holes for this purpose. The cam rod is then withdrawn, the fuel pins inserted, and the keys withdrawn. Hydraulic equipment for carrying out these operations for a large number of fuel rods simultaneously is also described.

  6. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Movafeghi, Amir

    2016-11-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  7. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    International Nuclear Information System (INIS)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser; Movafeghi, Amir

    2016-01-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  8. Prototypical spent nuclear fuel rod consolidation equipment: Phase 2, Final design report: Volume 4, Appendices: Part 3

    International Nuclear Information System (INIS)

    Ciez, A.P.

    1987-01-01

    The purpose of this manual is to provide assembly, installation, operation, maintenance, and off-normal recovery procedures for the Consolidation Equipment. The Consolidation System is a horizontal, dry system capable of processing one Pressurized Water Reactor (PWR) fuel assembly or one Boiling Water Reactor (BWR) fuel assembly at a time. The system will process all spent PWR and BWR fuels from the commercial US nuclear power reactor industry. Component changeouts for various fuel types have been minimized to reduce costs, required in-cell module storage space, and to increase efficiency by decreasing set-up time between fuel consolidation campaigns. The most important feature of the Westinghouse system is the ability to control the fuel rods at all times during the consolidation process from rod extraction, through canister loading. This features assures that the rods from two PWR fuel assemblies or four BWR fuel assemblies (minimum) can be loaded into one consolidated rods canister

  9. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  10. The thermo-mechanics of the PWR fuel rod; La thermomecanique du crayon de combustible REP

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France, EDF, 75 - Paris (France); Gautier, B.; Chaigne, G. [Electricite de France, Service Etudes et Projets Thermiques et Nucleaires, 75 - Paris (France)] [and others

    1999-03-29

    The fuel rod mechanics is of a great importance in the safety and performance of the reactors. In this domain a meeting has been organized by the SFEN the 18 march 1998 at Paris. With the participation of scientists from CEA, EDF and Framatome, the physics of the fuel rods was presented based on four main aspects. Two first papers dealt with the solicitations of the fuel rod in normal and accidental conditions. The physical phenomena under irradiation were then detailed in the four following talks. Three papers presented the simulation and the codes of the fuel-cladding interactions with the diabolo effect. The last paper was devoted to the experiment feedback and the research programs. (A.L.B.)

  11. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    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

  12. NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

    1998-08-01

    An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

  13. Fuel rod response to BWR power oscillations during anticipated transient without scram

    International Nuclear Information System (INIS)

    Cunningham, M.; Scott, H.

    1998-01-01

    The US NRC is examining fuel behaviour during a postulated BWR anticipated transient without scram (ATWS) with power oscillations to determine if current regulatory criteria are adequate. Currently, the 280 cal/g limit for RIAs is used to show that coolable geometry is maintained and pressure pulses are avoided during ATWSs. Two specific questions have now been raised about the continued use of the 280 cal/g value. First, this value was derived from energy deposition values whereas the regulatory requirements are written in terms of fuel enthalpy. The second is that fuel rod rupture with fuel dispersal has been observed in RIA tests with high bum-up fuel rods having energy deposition values well below the current limit. However, the BWR ATWS power oscillation transient is slower than a RIA power pulse, thus reducing the likelihood of failure. Therefore questions about the adequacy of the 280 cal/g limit do not necessarily imply unacceptable fuel damage occurring during such power oscillations and there is no immediate safety concern. The reported analysis, using the FRAPTRAN transient fuel rod analysis code, was thus undertaken to determine if further investigation might be appropriate and with the intention of starting some discussions about the issue. There was a comment that a limit of 100 cal/g fuel enthalpy had been mentioned following the scoping calculations but that perhaps enthalpy was not the main concern in an ATWS. It was also observed that cladding stresses are lower than in all RIA. The question was what really is the main concern. It was replied that the main concern was a question of maintaining a coolable geometry i.e. not loosing fuel particles out of the rod. And it was agreed that enthalpy may not be the important issue, rather that it previously had been used as the parameter and so had been considered. Confirmation of this presently being an evaluation and not a regulatory concern was sought and provided, it being pointed out that the NRC

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

    International Nuclear Information System (INIS)

    Dias, A.F.V.

    1980-10-01

    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.) [pt

  15. Burn up Theoretical Analysis of A Thorium Fuel Rod in Light Water Reactor

    International Nuclear Information System (INIS)

    Gaber, F.A.; Aziz, M.; Elsheikh, B.

    2008-01-01

    A computer model was designed to analyze the burn up and irradiation of both Th-Pu and Th-U fuel rod in a typical light water reactors conditions. MCNP computer model was designed to simulate the fuel rod burnup and evaluate neutron flux and group constants . A system of ordinary differential equations were solved numerically to evaluate the isotopic concentrations for both the two types of fuel using the previous calculated data from MCNP model. The results are analyzed and compared with published data where satisfactory agreement was found

  16. Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

    International Nuclear Information System (INIS)

    Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae

    2016-01-01

    The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods

  17. Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods.

  18. The light-water-reactor version of the Uranus integral fuel-rod code

    International Nuclear Information System (INIS)

    Moreno, A.; Lassmann, K.

    1977-01-01

    The LWR of the Uranus code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via Uranus results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analysis of in-pile experiments, they illustrate rather typical and diverse Uranus capabilities. The performance test shows that Uranus is reliable and efficient, thus the code is a most valuable tool in fuel fod analysis work. K. Lassmann developed the LWR version of the Uranus code, material properties were reviewed and supplied by A. Moreno. (author)

  19. Thermal performance of annular-coated and sphere-pac LWR fuel rod designs

    International Nuclear Information System (INIS)

    Guenther, R.J.; Hsieh, K.A.; Barner, J.O.; Freshley, M.D.

    1980-01-01

    Two FCI-resistant UO 2 fuel rod designs are being compared to a reference design in irradiation tests in the Halden Boiling Water Reactor (HBWR) as part of the DOE-sponsored Fuel Performance Improvement Program (FPIP). The primary fuel design (annular-coated-pressurized) incorporates annular pellets, a graphite coating on the inner surface of the Zircaloy cladding, and pressurized helium fill gas. Also being investigated is an 87% smear density sphere-pac design with pressurized helium fill gas. The solid pellet (reference) and annular-coated designs described had helium fill gas at approx. 100 kPa and the sphere-pac rods were pressurized at approx. 455 kPa

  20. Development of techniques for joining fuel rod simulators to test assemblies

    International Nuclear Information System (INIS)

    Moorhead, A.J.; Reed, R.W.

    1980-01-01

    A unique tubular electrode carrier is described for gas tungsten-arc welding small-diameter nuclear fuel rod simulators to the tubesheet of a test assembly. Both the close-packed geometry of the array of simulators and the extension of coaxial electrical conductors from each simulator hindered access to the weld joint. Consequently, a conventional gas tungsten-arc torch could not be used. Two seven-rod assemblies that were mockups of the simulator-to-tubesheet joint area were welded and successfully tested. Modified versions of the electrode carrier for brazing electrical leads to the upper ends of the fuel pin simulators are also described. Satisfactory brazes have been made on both single-rod mockups and an array of 25 simulators by using the modified electrode carrier and a filler metal with a composition of 71.5 Ag-28 Cu-0.5 Ni

  1. Comparison between temperature distributions of an annular fuel rod of circular cross-section and of a hemoglobin shaped cross-section rod for PWR reactors in steady state conditions

    International Nuclear Information System (INIS)

    Oliveira, Maria Vitória A. de; Alvim, Antônio Carlos Marques

    2017-01-01

    The objective of this work is to make a comparison between the temperature distributions of an annular fuel rod of circular cross-section and a hemoglobin shaped cross-section for PWR reactors in steady state conditions. The motivation for this article is due to the fact that the symmetric form of the red globules particles allows the O 2 gases to penetrate the center of the cell homogeneously and quickly. The diffusion equation of gases in any environment is very similar to the heat diffusion equation: Diffusion - Fick's Law; Heat Flow - Fourier; where, the temperature (T) replaces the concentration (c). In previous works the comparison between the shape of solid fuel rods with circular section, and a with hemoglobin-shaped cross-section has proved that this new format optimizes the heat transfer, decreasing the thermal resistance between the center of the UO 2 pellets and the clad. With this, a significant increase in the specific power of the reactor was made possible (more precisely a 23% increase). Currently, the advantages of annular fuel rods are being studied and recent works have shown that 12 x 12 arrays of annular fuel rods perform better, increasing the specific power of the reactor by at least 20% in relation to solid fuel rods, without affecting the safety of the reactor. Our proposal is analyzing the temperature distribution in annular fuel rods with cross sections with red blood cell shape and compare with the theoretical results of the annular fuel rods of circular cross section, initially in steady state. (author)

  2. Comparison between temperature distributions of an annular fuel rod of circular cross-section and of a hemoglobin shaped cross-section rod for PWR reactors in steady state conditions

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Maria Vitória A. de; Alvim, Antônio Carlos Marques, E-mail: moliveira@con.ufrj.br, E-mail: alvim@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    The objective of this work is to make a comparison between the temperature distributions of an annular fuel rod of circular cross-section and a hemoglobin shaped cross-section for PWR reactors in steady state conditions. The motivation for this article is due to the fact that the symmetric form of the red globules particles allows the O{sub 2} gases to penetrate the center of the cell homogeneously and quickly. The diffusion equation of gases in any environment is very similar to the heat diffusion equation: Diffusion - Fick's Law; Heat Flow - Fourier; where, the temperature (T) replaces the concentration (c). In previous works the comparison between the shape of solid fuel rods with circular section, and a with hemoglobin-shaped cross-section has proved that this new format optimizes the heat transfer, decreasing the thermal resistance between the center of the UO{sub 2} pellets and the clad. With this, a significant increase in the specific power of the reactor was made possible (more precisely a 23% increase). Currently, the advantages of annular fuel rods are being studied and recent works have shown that 12 x 12 arrays of annular fuel rods perform better, increasing the specific power of the reactor by at least 20% in relation to solid fuel rods, without affecting the safety of the reactor. Our proposal is analyzing the temperature distribution in annular fuel rods with cross sections with red blood cell shape and compare with the theoretical results of the annular fuel rods of circular cross section, initially in steady state. (author)

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

    International Nuclear Information System (INIS)

    Gill, C.R.; Coddington, P.

    1988-05-01

    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)

  4. Comparison with experiment of COMETHE III-L fuel rod behaviour predictions

    International Nuclear Information System (INIS)

    Vliet, J. van; Billaux, M.

    1983-01-01

    A comparison is presented between experimental results and COMETHE III-L fuel rod behaviour predictions. The first part of the paper focuses on mechanical aspects, with as main experiments, AECL X-264 and Studsvik Interramp. The second part presents the results of a wide FGR benchmarking campaign, with a reference to previous COMETHE versions. It appears that the variance between experiment and calculation has decreased by a factor four when the III-J version was improved into the III-L version. As conclusion, some COMETHE III-L calculations are presented in order to illustrate its capability of predicting fuel rod performance limits. (author)

  5. FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Sakagami, M.

    1984-01-01

    FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

  6. Analyses of expected rod performance during the dry storage of spent fuel

    International Nuclear Information System (INIS)

    Einziger, R.E.

    1982-08-01

    Within the next ten years, a number of utilities will be forced to increase their interim spent-fuel-storage capability or face the loss of full-core reserve. Dry storage is being considered to fill this need. This paper analyzes the fuel-rod-performance data supporting dry storage and discusses areas where there are still outstanding questions. Three storage temperature ranges (T 0 C, 250 0 C 0 C and T > 400 0 C), two atmospheres (inert, unlimited air) and two initial fuel-rod conditions (intact, breached) are considered. It is concluded that a fuel-performance data base exists that indicates that storage below 250 0 C can be accomplished with long-term fuel pellet and cladding stability. At higher temperatures, analytic studies and laboratory experiments are needed especially to extrapolate and interpret the result of demonstration tests. 2 figures, 2 tables

  7. Prediction of failure enthalpy and reliability of irradiated fuel rod under reactivity-initiated accidents by means of statistical approach

    International Nuclear Information System (INIS)

    Nam, Cheol; Choi, Byeong Kwon; Jeong, Yong Hwan; Jung, Youn Ho

    2001-01-01

    During the last decade, the failure behavior of high-burnup fuel rods under RIA has been an extensive concern since observations of fuel rod failures at low enthalpy. Of great importance is placed on failure prediction of fuel rod in the point of licensing criteria and safety in extending burnup achievement. To address the issue, a statistics-based methodology is introduced to predict failure probability of irradiated fuel rods. Based on RIA simulation results in literature, a failure enthalpy correlation for irradiated fuel rod is constructed as a function of oxide thickness, fuel burnup, and pulse width. From the failure enthalpy correlation, a single damage parameter, equivalent enthalpy, is defined to reflect the effects of the three primary factors as well as peak fuel enthalpy. Moreover, the failure distribution function with equivalent enthalpy is derived, applying a two-parameter Weibull statistical model. Using these equations, the sensitivity analysis is carried out to estimate the effects of burnup, corrosion, peak fuel enthalpy, pulse width and cladding materials used

  8. Grid for nuclear fuel rod assembly

    International Nuclear Information System (INIS)

    Brayman, K.W.; George, D.K.; Rawlings, J.C.; Dix, G.E.

    1976-01-01

    A grid is described for placing a least four corner fuel rods in a tubular flow channel of a nuclear reactor. It includes a bearer component composed of four side strips joined by four corner strips so as to form a rigid unit structure, each side strip having an L-shaped piece adjacent at each of its ends to a lug of each L-shaped piece extending to the adjacent end of its associated side strip [fr

  9. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  10. Computer analysis of elongation of the WWER fuel rod claddings

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2008-01-01

    In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

  11. Numerically predicting horizontally oriented spent fuel rod surface temperatures

    International Nuclear Information System (INIS)

    Wix, S.D.; Koski, J.A.

    1993-01-01

    A comparison between numerical calculations with use of commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degrees C and 23 degrees C for the low heat dissipation and high dissipation, respectively. The temperature predictions using helium as a fill gas are lower than the experimental data for the low and medium heat dissipation levels. The temperature predictions are 1 degrees C and 6 degrees C lower than the experimental data for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degrees C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include a experimental uncertainity in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This works demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects, such as axial heat transfer through the spent fuel rods, will be increasingly important as the amount of dissipated heat increases

  12. Numerically predicting horizontally oriented spent fuel rod surface temperatures

    International Nuclear Information System (INIS)

    Wix, S.D.; Koski, J.A.

    1992-01-01

    A comparison between numerical calculations with use of commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degree C and 23 degree C for the low heat dissipation and high heat dissipation, respectively. The temperature predictions using helium as a fill gas are lower than the experimental data for the low and medium heat dissipation levels. The temperature predictions are 1 degree C and 6 degree C lower than the experimental data for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degree C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include experimental uncertainty in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This work demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects, such as axial heat transfer through the spent fuel rods, will be increasingly important as the amount of dissipated heat increases

  13. Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

    International Nuclear Information System (INIS)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

    1995-01-01

    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 B 4 C absorber rod and the stainless steel grid spacers on the 'low-temperature' bundle damage initiation and progression. The B 4 C 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 occured. (orig./HP)

  14. Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

    1995-08-01

    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 B{sub 4}C absorber rod and the stainless steel grid spacers on the `low-temperature` bundle damage initiation and progression. The B{sub 4}C 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 occured. (orig./HP)

  15. Thermal performance of the nuclear fuel rods submitted to angular variation of the heat exchanger coefficients

    International Nuclear Information System (INIS)

    Carvalho, A.M.M. de.

    1984-01-01

    Generally, LMFBR fuel rods consist of fuel pellets encapsulated in cladding tubes. These tubes are wrapped by a helical wire, working as a spacer. Distortions in the rod temperature distribution and in the external heat flux can be generated by angular variations in the local heat transfer coefficients due to the wire, by excentricity between pellet and clad or by ovalization of the cladding tube. Also, the temperature distributions can be affected by fuel densification, reestructuring and swelling. The present work consists of the development of a computer code in order to analyse the fuel rod performance as function of geometrical and operational effects, in steady state regime. (Author) [pt

  16. Evaluation of the effect of probe design parameters on ECT signal and development of eddy current probe for irradiated fuel rods

    International Nuclear Information System (INIS)

    Kwank, S. W.; Han, Y. K.; Woo, S. K.; Kim, T. W.; Park, J. Y.; Kim, B. J.; Park, J. Y.

    1999-01-01

    Eddy current test(ECT) is used to inspect not only the failed fuel rods but also peripheral rods during repairing of the failed fuel rods, to detect internal defects in irradiated fuel rods which could not be detected by ultrasonic test and visual test, and to obtain the data for determining the root cause of fuel rod failure. This study evaluates the effect of properties of test article, irradiated fuel rods, on the impedance diagram in order to reduce the difficulty of ECT signal analysis. The optimum eddy current probe design conditions for inspecting the irradiated fuel rods, is estimate by using experimental equations and the probe is manufactured based on the estimated conditions. The performance of developed eddy current probe and the optimum conditions is proved through characteristic comparison experiment with the probe purchased from the foreign vendor

  17. The ''THERMOST'' for analysing thermo-structural behaviour of LWR fuel rod under PCI conditions

    International Nuclear Information System (INIS)

    Nuno, H.; Ogawa, S.; Kobayashi, H.

    1983-01-01

    As one of the methods for evaluating the fuel rod performances under power ramping or load following operations, the combined ''FROST'' and ''THERMOST'' system has been developed and being brought into practical use. The former had already been presented at Blackpool Meeting in 1978, and the latter is going to be presented in this paper. The major purpose of the THERMOST is to analyse very detailed thermal and structural fuel behaviours in a rather localized part of fuel rod whereas the FROST deals with whole-rod-wide general performances. The code handles 2-dimensional thermal and structural analyses simultaneously by using finite element method, in axial section wide or in lateral section wide. It consists of a fundamental FEM system of generalized constitution and its surrounding subroutine system which characterizes fuel behaviours such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer elements (6 kinds) and structural analysis by axisymmetric ring and lateral plane elements (6 kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping condition is presented with some inpile test data. (author)

  18. 3D finite element analysis of a nuclear fuel rod with gap elements between the pellet and the cladding

    International Nuclear Information System (INIS)

    Kang, Chang-Hak; Lee, Sung-Uk; Yang, Dong-Yol; Kim, Hyo-Chan; Yang, Yong-Sik

    2016-01-01

    Nuclear fuel rods which comprises an important component of a nuclear power plant are composed of nuclear fuel and cladding. Simulating the nuclear fuel rod using a computer program is the universal method to verify its safety. The computer program used for this is called the fuel performance code. The main objective of this study is to simulate the nuclear fuel rod behavior considering the gap conductance using three-dimensional gap elements. Gap elements are used because, unlike other methods, this approach does not require special methods or other variables such as the Lagrange multiplier. In this work, a nuclear fuel rod has been simulated and the results are compared with the experimental results. (author)

  19. Models of multi-rod code FRETA-B for transient fuel behavior analysis

    International Nuclear Information System (INIS)

    Uchida, Masaaki; Otsubo, Naoaki.

    1984-11-01

    This paper is a final report of the development of FRETA-B code, which analyzes the LWR fuel behavior during accidents, particularly the Loss-of-Coolant Accident (LOCA). The very high temperature induced by a LOCA causes oxidation of the cladding by steam and, as a combined effect with low external pressure, extensive swelling of the cladding. The latter may reach a level that the rods block the coolant channel. To analyze these phenomena, single-rod model is insufficient; FRETA-B has a capability to handle multiple fuel rods in a bundle simultaneously, including the interaction between them. In the development work, therefore, efforts were made for avoiding the excessive increase of calculation time and core memory requirement. Because of the strong dependency of the in-LOCA fuel behavior on the coolant state, FRETA-B has emphasis on heat transfer to the coolant as well as the cladding deformation. In the final version, a capability was added to analyze the fuel behavior under reflooding using empirical models. The present report describes the basic models of FRETA-B, and also gives its input manual in the appendix. (author)

  20. Corrosion performance of optimised and advanced fuel rod cladding in PWRs at high burnups

    International Nuclear Information System (INIS)

    Jourdain, P.; Hallstadius, L.; Pati, S.R.; Smith, G.P.; Garde, A.M.

    1997-01-01

    The corrosion behaviour both in-pile and out-of-pile for a number of cladding alloys developed by ABB to meet the current and future needs for fuel rod cladding with improved corrosion resistance is presented. The cladding materials include: 1) Zircaloy-4 (OPTIN) with optimised composition and processing and Zircaloy-2 optimised for Pressurised Water Reactors (PWR), (Zircaloy-2P), and 2) several alternative zirconium-based alloys with compositions outside the composition range for Zircaloys. The data presented originate from fuel rods irradiated in six PWRs to burnups up to about 66 MWd/kgU and from tests conducted in 360 o water autoclave. Also included are in-pile fuel rod growth measurements on some of the alloys. (UK)

  1. Fuel rod D07/B15 from Ringhals 2 PWR: Source material for corrosion/leach tests in groundwater. Fuel rod/pellet characterization program. Pt. 1

    International Nuclear Information System (INIS)

    Forsyth, R.

    1987-03-01

    A joint SKB/STUDSVIK experimental program to determine the corrosion rates and to establish the corrosion mechanisms of spent UO 2 fuel in groundwater under both oxidizing and reducing conditions is in progress in the Hot Cell Laboratory of Studsvik Energiteknik AB. High burnup fuel of both BWR and PWR type are studied. Characterization of the spent fuel at both rod and pellet level is an important part of the experimental program. Experiments on PWR fuel have been concentrated so far on specimens from one rod, manufacturer's number 03688, which had occupied position B15 in assembly D07. This assembly had been irradiated for 5 cycles in the Ringhals 2 reactor between 1977 and 1983. The calculated assembly burnup was 41.3 MWd/kg U. The present report is a collection of separate reports describing those items in the characterization program which have been performed so far. No overall summary of the experimental results is given here, and the report should be viewed as a collection of reference data. (orig.)

  2. Calculation of fission gases internal pressure in nuclear fuel rods

    International Nuclear Information System (INIS)

    Vasconcelos Santana, M. de.

    1981-12-01

    Models concerning the principal phenomena, particularly thermal expansion, fuel swelling, densification, reestructuring, relocation, mechanical strain, fission gas production and release, direct or indirectly important to calculate the internal pressure in nuclear fuel rods were analysed and selected. Through these analyses a computer code was developed to calculate fuel pin internal pressure evolution. Three different models were utilized to calculate the internal pressure in order to select the best and the most conservative estimate. (Author) [pt

  3. Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo; Sobajima, Makoto.

    1993-09-01

    This report presents experimental data for Test TS-3 which was the third 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 September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) 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 26 Gwd/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 energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. 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)

  4. Removal and replacement of fuel rods in nuclear fuel assembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Ferlan, S.J.

    1983-01-01

    Apparatus for replacing components of a nuclear fuel assembly stored in a pit under about 10 m. of water. The fuel assembly is secured in a container which is rotatable from the upright position to an inverted position in which the bottom nozzle is upward. The bottom nozzle plate is disconnected from the control-rod thimbles by means of a cutter for severing the welds. To guide and provide lateral support for the cutter a fixture including bushings is provided, each encircling a screw fastener and sealing the region around a screw fastener to trap the chips from the severed weld. Chips adhering to the cutter are removed by a suction tube of an eductor. (author)

  5. Equipment to weld fuel rods of mixed oxides

    International Nuclear Information System (INIS)

    Aparicio, G.; Orlando, O.S.; Olano, V.R.; Toubes, B.; Munoz, C.A.

    1987-01-01

    Two welding outfits system T1G were designed and constructed to weld fuel rods with mixed oxides pellets (uranium and plutonium). One of them is connected to a glove box where the loading of sheaths takes place. The sheaths are driven to the welder through a removable plug pusher in the welding chamber. This equipment was designed to perform welding tests changing the parameters (gas composition and pressure, welding current, electrode position, etc.). The components of the welder, such as plug holder, chamber closure and peripheral accessories, were designed and constructed taking into account the working pressures in the machine, which is placed in a controlled area and connected to a glove box, where special safety conditions are necessary. The equipment to weld fuel bars is complemented by another machine, located in cold area, of the type presently used in the fuel elements factory. This equipment has been designed to perform some welding operations in sheaths and mixed oxide rods of the type Atucha I and II. Both machines have a programmed power supply of wide range and a vacuum, and pressurizing system that allows the change of parameters. Both systems have special features of handling and operation. (Author)

  6. Cladding temperature measurement by thermocouples at preirradiated LWR fuel rod samples

    International Nuclear Information System (INIS)

    Leiling, W.

    1981-12-01

    This report describes the technique to measure cladding temperatures of test fuel rod samples, applied during the in-pile tests on fuel rod failure in the steam loop of the FR2 reactor. NiCr/Ni thermocouples with stainless steel and Inconel sheaths, respectively,of 1 mm diameter were resistance spot weld to the outside of the fuel rod cladding. For the pre-irradiated test specimens, welding had to be done under hot-cell conditions, i.e. under remote handling. In order to prevent the formation of eutectics between zirconium and the chemical elements of the thermocouple sheath at elevated temperatures, the thermocouples were covered with a platinum jacket of 1.4 mm outside diameter swaged onto the sheath in the area of the measuring junction. This thermocouple design has worked satisfactorily in the in-pile experiments performed in a steam atmosphere. Even in the heatup phase, in which cladding temperatures up to 1050 0 C were reached, only very few failures occured. This good performance is to a great part due to a careful control and a thorough inspection of the thermocouples. (orig.) [de

  7. Performance of artificially defected LWR fuel rods in an unlimited air dry storage atmosphere

    International Nuclear Information System (INIS)

    Einziger, R.E.; Knecht, R.L.; Cantley, D.A.; Cook, J.A.

    1983-09-01

    Thus far the tests are inconclusive as to whether breached LWR fuel can be stored at 230 0 C for long periods of time in air without fuel oxidation and dispersion. There is every indication, as expected, that there is no oxidation problem in an inert atmosphere. Only one of four defects exposed to unlimited air gave any indication of fuel oxidation. It has been suggested that this might be an incubation effect and continued operation would result in oxidation occurring at all four defects. As yet the destructive examination of the BWR rod has not been completed, so it is not possible to determine if cladding splitting was due to an anomoly in this test rod or something that can be expected in LWR rods in general. Thus far there is no indication of respirable particle dispersal even if fuel oxidation does occur

  8. The use of eddy current testing for nuclear fuel rods cladding evaluation

    International Nuclear Information System (INIS)

    Silva Junior, Silverio F. da; Alencar, Donizete A.; Brito, Mucio Jose D. de

    2007-01-01

    Nuclear fuel rods cladding must be tested after their manufacture and during their operational life. This paper describes a study about the use of eddy current test method as a nondestructive tool for nuclear fuel rods cladding evaluation. The experiments were carried out using two different probes: an external probe and an internal probe. The main goal was to verify the sensitivity of the eddy current test system, to develop calibration and reference standards and to establish the main capabilities and limitations presented by this test method for this application. (author)

  9. Mechanical fragmentation of nuclear reactor fuel assemblies by the double cutting method

    International Nuclear Information System (INIS)

    Voitsekhovskii, B.V.; Istomin, V.L.; Mitrofanov, V.V.

    1995-01-01

    A method is described for cutting a spent fuel assembly with straight shears into pieces of a prescribed size. The method does not require separation of the casing and the lattices. The double cutting method is briefly described, and experiments designed for cutting BN-350 and VVER-440 fuel assemblies are outlined. The testing showed that the cutting method was suitable for mechanical polarization of fuel assemblies. The investigations led to the development of turnkey industrial equipment for cutting spent fuel assemblies of different geometries with a maximum size up to 170 mm. 6 refs., 8 figs., 1 tab

  10. Apparatus and method for preventing the rotation of rods used in nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Pilgrim, L.G. Jr.; Jackson, L.F.

    1985-01-01

    Apparatus and method for preventing the rotation of one or more elongated rods used in nuclear fuel assemblies include an end plug secured to one longitudinal end of such an elongated rod and having an out-of-cavity, non-round structure affixed thereto and configured to mate with a complementary shaped structure in a lower tie plate of a nuclear fuel assembly in such a manner as to prevent the rotation of the rod about its longitudinal axis. In one embodiment, the end plug includes a pair of flats formed on a portion of the end plug and configured to abut against a pair of flats formed on the outer surface of a cylindrical boss or sleeve of the lower tie plate, thereby to prevent the rotation of the rod. In another embodiment, four grooves, disposed 90 0 apart about the periphery of an end plug of a rod form a spline. The grooves are configured to receive four, radially inwardly protruding, key members disposed 90 0 apart about the periphery of a sleeve secured to the lower tie plate, thereby to prevent the rotation of the rod. In a further embodiment, a sleeve is secured to an end plug of a rod and includes four elongated slots disposed 90 0 apart about the periphery of the sleeve and configured in width, depth and spacing to receive and mate with four web portions of the lower tie plate of the nuclear fuel assembly, thereby to secure the rod against rotation about its longitudinal axis

  11. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, J.M.

    1980-01-01

    A control algorithm has been derived for an HTGR Fuel Rod Fabrication Process utilizing the method of G.E.P. Box and G.M. 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 ref

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

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.; Passage, G.; Stefanova, S.

    2009-01-01

    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

  13. 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)

    Pantoja C, R.

    2010-01-01

    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

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

    International Nuclear Information System (INIS)

    Villarino, E.A.

    1990-01-01

    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) [es

  15. Heat removal in gas-cooled fuel rod clusters

    International Nuclear Information System (INIS)

    Rehme, K.

    1975-01-01

    For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

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

    International Nuclear Information System (INIS)

    Rosa, J.M.

    1986-01-01

    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

  17. SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

    International Nuclear Information System (INIS)

    Gulden, W.

    1977-08-01

    The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

  18. 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.

  19. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Joo, K. N.; Park, S. J.; Kang, Y. H.; Kim, Y. K.; Yeum, K. I. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. Therefore, the out of pile test system for pressure measurement was developed, and the test with the LVDT at room temperature were performed. This test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2}, and repeated 6 times at same condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule.

  20. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Min, Sohn Jae; Kang, Y. H.; Kim, B. G. [and others

    2001-11-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO, the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT. The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. The out-of-pile test system for pressure measurement was developed, and the test with the LVDT at room temperature(19 .deg. C) were performed. A out-of-pile test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2} and repeated 6 times at each condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. This report describes the system configuration, the out-of-pile test procedures, and the results. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics for the detail design of the fuel irradiation capsule.

  1. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    International Nuclear Information System (INIS)

    Sengpiel, W.

    1980-12-01

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW) [de

  2. Design criteria for confidence in the manufacture of BWR fuel rods

    International Nuclear Information System (INIS)

    Anantharaman, K.; Basu, S.; Anand, A.K.; Mehta, S.K.

    Based on the experience of fuel manufacture for BWR type reactors in India, the parameters which need stringent quality control, are discussed. The design specifications of the fuel rods as well as the cladding material and tubes are reported. The defect mechanisms to be taken into account and the fuel failure in reference to the variation of mechanical properties of the cladding are also described. (K.B.)

  3. FEMAXI-III: a computer code for the analysis of thermal and mechanical behavior of fuel rods

    International Nuclear Information System (INIS)

    Nakajima, Tetsuo; Ichikawa, Michio; Iwano, Yoshihiko; Ito, Kenichi; Saito, Hiroaki; Kashima, Koichi; Kinoshita, Motoyasu; Okubo, Tadatsune.

    1985-12-01

    FEMAXI-III is a computer code to predict the thermal and mechanical behavior of a light water fuel rod during its irradiation life. It can analyze the integral behavior of a whole fuel rod throughout its life, as well as the localized behavior of a small part of fuel rod. The localized mechanical behavior such as the cladding ridge deformation is analyzed by the two-dimensional axisymmetric finite element method. FEMAXI-III calculates, in particular, the temperature distribution, the radial deformation, the fission gas release, and the inner gas pressure as a function of irradiation time and axial position, and the stresses and strains in the fuel and cladding at a small part of fuel rod as a function of irradiation time. For this purpose, Elasto-plasticity, creep, thermal expansion, fuel cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, and fuel-cladding mechanical interaction are modelled and their interconnected effects are considered in the code. Efforts have been made to improve the accuracy and stability of finite element solution and to minimize the computer memory and running time. This report describes the outline of the code and the basic models involved, and also includes the application of the code and its input manual. (author)

  4. Performance evaluation of the Loviisa advanced type fuel rods

    International Nuclear Information System (INIS)

    Ranta-Puska, K.; Pihlatie, M.

    2001-01-01

    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 85 Kr 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

  5. Proceedings of the specialist meeting on nuclear fuel and control rods: operating experience, design evolution and safety aspects

    International Nuclear Information System (INIS)

    1997-01-01

    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)

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

    International Nuclear Information System (INIS)

    Sepold, L.

    1995-06-01

    The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B 4 C/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 B 4 C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B 4 C 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.) [de

  7. Conservative performance analysis of a PWR nuclear fuel rod using the FRAPCON code

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Fabio Branco Vaz de; Sabundjian, Gaiane, E-mail: fabio@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In this paper, some of the preliminary results of the sensitivity and conservative analysis of a hypothetical pressurized water reactor fuel rod are presented, using the FRAPCON code as a basic and preparation tool for the future transient analysis, which will be carried out by the FRAPTRAN code. Emphasis is given to the evaluation of the cladding behavior, since it is one of the critical containment barriers of the fission products, generated during fuel irradiation. Sensitivity analyses were performed by the variation of the values of some parameters, which were mainly related with thermal cycle conditions, and taking into account an intermediate value between the realistic and conservative conditions for the linear heat generation rate parameter, given in literature. Time lengths were taken from typical nuclear power plant operational cycle, adjusted to the obtention of a chosen burnup. Curves of fuel and cladding temperatures, and also for their mechanical and oxidation behavior, as a function of the reactor operation's time, are presented for each one of the nodes considered, over the nuclear fuel rod. Analyzing the curves, it was possible to observe the influence of the thermal cycle on the fuel rod performance, in this preliminary step for the accident/transient analysis. (author)

  8. Behavior of defective LWR-type fuel rods irradiated under postulated accident conditions

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Croucher, D.W.; Seiffert, S.L.; Cook, B.A.; Kerwin, D.K.; Mehner, A.S.; Ploger, S.A.

    1979-05-01

    The irradiation experiments reported here have been conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., for the United States Nuclear Regulatory Commission in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Five of the rods were irradiated in PCM tests and one in a LOC test. During these tests, the six rods lost cladding integrity prior to or during the transient phase of the test due to either manufacturing defects or intentional rod design and operation. Of the five defective rods tested under PCM conditions, one (Rod IE-008, Test IE-1) had a hydride rupture below the region of the rod, which was in film boiling during the transient; two (Rod A-0021, Test PCM-3 and Rod IE-019, Test IE-5) contained defects (a pin hole and a small axial crack, respectively) within the film boiling zone; and two (Rod 201-1, Test PCM-1 and Rod 205-8, Test PCM-5) failed by cladding embrittlement within the film boiling zone. Rod 312-3 was waterlogged before being subjected to LOC conditions in Test LLR-3

  9. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

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

    International Nuclear Information System (INIS)

    1986-01-01

    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

  11. Multiple fuel rod gripper

    International Nuclear Information System (INIS)

    Shields, E.P.

    1987-01-01

    An apparatus is described for gripping an array of rods comprising: (a) gripping members grippingly engageable with the rods, each of which has a hollow portion terminating in an open end for receiving the end of one of the rods; (b) a closing means for causing the hollow portion of each of the gripping members to apply substantially the same gripping force onto the end of its respective rod, including (i) a locking plate having a plurality of tapered holes registrable with the array of rods, wherein the exterior of each of the gripping members is tapered and nested within one of the tapered holes, (ii) a withdrawing means having a hydraulic plunger operatively connected to each of the gripping members for applying a substantially identical withdrawing force on each of the gripping members, whereby the hollow portion of each of the gripping members applies substantially the same gripping force on its respective rod, and (c) means for detecting whether each of the gripping members has grippingly engaged its respective rod

  12. Fuel rod with axial regions of annular and standard fuel pellets

    International Nuclear Information System (INIS)

    Freeman, T.R.

    1991-01-01

    This patent describes a fuel rod for use in a nuclear reactor fuel assembly. It comprises: an elongated hollow cladding tube; a pair of end plugs connected to and sealing the cladding tube at opposite ends of thereof; and an axial stack of fuel pellets contained in and extending between the end plugs at the opposite ends of the tube, all of the fuel pellets contained in the tube being composed of fissile material being enriched above the level of natural enrichment; the fuel pellets in the stack thereof being provided in an arrangement of axial regions. The arrangement of axial regions including a pair of first axial regions defined respectively at the opposite ends of the pellet stack adjacent to the respective end plugs. The pellets in the first axial regions being identical in number and having annular configurations with an annulus of a first void size. The arrangement of axial regions also including another axial region defined between the first axial regions, some of the pellets in the another axial region having solid configurations

  13. Status of rod consolidation

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1985-04-01

    Two of the factors that need to be taken into account with rod consolidation are (1) the effects on rods from their removal from the fuel assembly and (2) the effects on rods as a result of the consolidation process. Potential components of both factors are described in the report. Discussed under (1) are scratches on the fuel rod surfaces, rod breakage, crud, extended burnup, and possible cladding embrittlement due to hydrogen injection at BWRs. Discussed under (2) are the increased water temperature (less than 10 0 C) because of closer packing of the rods, formation of crevices between rods in the close-packed mode, contact with dissimilar metals, and the potential for rapid heating of fuel rods following the loss of water from a spent fuel storage pool. Another factor that plays an important role in rod consolidation is the cost of disposal of the nonfuel-bearing components of the fuel assembly. Also, the dose rate from the components - especially Inconel spacer grids - can affect the handling procedures. Several licensing issues that exist are described. A list of recommendations is provided. 98 refs., 5 figs., 5 tabs

  14. Impact of UO{sub 2} Enrichment of Fuel Zoning Rods in Long Cycle Operation of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ho Cheol; Lee, Deokjung [KHNP CRI, Daejeon (Korea, Republic of); Jeong, Eun; Choe, Jiwon [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    Extending the cycle length can not only increase the energy production, but also bring down outage costs by reducing the number of refueling outages during the lifetime of a nuclear power plant. It is reasonable that more fresh fuels are loaded for long cycle operation. However, minimizing the number of fresh fuels is essential in aspect of fuel economics. This can cause high power peaking near the water holes, due to increased thermalization of neutrons in those regions. To prevent this, special fuel zoning rods are used and surround the water holes. These rods use lower-enriched uranium (they have an enrichment rate lower than the other fuel rods). If we adjust the enrichment rate of fuel zoning rods, we can reduce power peaking and moreover increase cycle length. In this paper, we designed a core suitable for long cycle operation and we conducted sensitivity tests of fuel cycle length on UO2 enrichment rate in fuel zoning region in order to extend the cycle length while using the same number of fresh fuels. The correlations between the fuel zoning enrichment and cycle length, peaking factor, CBC and shutdown margin were analyzed. The more the enrichment rate in fuel zoning region increases, the more the fuel cycle length increases. At the same time, CBC, Fq and shutdown margin do not change significantly. Increasing the fuel zoning enrichment rate presents the right property of increasing the fuel cycle length without causing a large change to CBC, Fq and shutdown margin. In conclusion, by increasing the uranium enrichment rate in fuel zoning region, fuel cycle length can be increased and the safety margins can be maintained for long cycle operation of cores.

  15. Investigation of 3H and 14C inventory and distribution in spent BWR fuel rods

    International Nuclear Information System (INIS)

    Bleier, A.; Beuerle, M.; Neeb, K.H.

    1984-10-01

    In order to obtain reliable data for fuel reprocessing and waste disposal, the T and C-14 inventory, distribution and behaviour was investigated on a typical LWR fuel rod discharged from a BWR plant. The results showed that 50 ± 5% of the T generated in the fuel is present in the cladding after reactor operation. The remainder of the T stays with the fuel. Related to the reactor power the total T inventory corresponds to a T production rate of 19 000 Ci/GW e . a. The C-14 built up in the fuel represents approximately 60% of the C-14 inventory of the BWR fuel rod. The remaining part of C-14 (about 40%) experimentally determined by this analysis for the first time is generated in the cladding. From the total C-14 inventory a C-14 production rate of 17,5 Ci/GW e . a can be calculated. The fill gas contains only negligible fractions of both nuclides. The results obtained in this program are generally in good agreement with the data of theoretical estimates and with results of earlier investigations on PWR fuel rods. (orig.) [de

  16. Assessment of the prediction capability of the TRANSURANUS fuel performance code on the basis of power ramp tested LWR fuel rods

    International Nuclear Information System (INIS)

    Pastore, G.; Botazzoli, P.; Di Marcello, V.; Luzzi, L.

    2009-01-01

    The present work is aimed at assessing the prediction capability of the TRANSURANUS code for the performance analysis of LWR fuel rods under power ramp conditions. The analysis refers to all the power ramp tested fuel rods belonging to the Studsvik PWR Super-Ramp and BWR Inter-Ramp Irradiation Projects, and is focused on some integral quantities (i.e., burn-up, fission gas release, cladding creep-down and failure due to pellet cladding interaction) through a systematic comparison between the code predictions and the experimental data. To this end, a suitable setup of the code is established on the basis of previous works. Besides, with reference to literature indications, a sensitivity study is carried out, which considers the 'ITU model' for fission gas burst release and modifications in the treatment of the fuel solid swelling and the cladding stress corrosion cracking. The performed analyses allow to individuate some issues, which could be useful for the future development of the code. Keywords: Light Water Reactors, Fuel Rod Performance, Power Ramps, Fission Gas Burst Release, Fuel Swelling, Pellet Cladding Interaction, Stress Corrosion Cracking

  17. Grab structure of a lifting structure in particular for use in a nuclear reactor for lifting and lowering of fuel elements and fuel rods

    International Nuclear Information System (INIS)

    Dose, G.

    1979-01-01

    A guide tower projects perpendicularly downward from the carriage of the charging machine. It can be rotated about its perpendicular axis. The tower is used to displace a hollow grab structure with two grabs. They can be opened and closed, the closed position being retained as long as they carry the fuel elements or rods. The power and interlocking equipment is installed one unit above the other in the joint grab housing. The tower with the integrated fuel element grab and the rod grab is rotated about its perpendicular axis for inspection of the fuel elements or rods. (DG) [de

  18. Post-irradiation examination of Al-61 wt% U3Si fuel rods from the NRU reactor

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    1997-01-01

    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 D 2 O 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 % 235 U 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% U 3 Si fuel irradiated in the NRU reactor. (author)

  19. Procedure and apparatus for measuring the radial gap between fuel and surrounding cladding in a fuel rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Olshausen, K.D.

    1976-01-01

    A device is described for measuring non-destructively the annular fuel-cladding gap in an irradiated or fresh fuel rod. The principle applied is that a force is applied to an arm which presses the cladding diametrically, thus deforming it until it touches the fuel pellet. By presenting the values of the force applied and the deformation produced on an XY recorder, the width of the gap is obtained. Alternatively the gap width may be obtained digitally. Since the gap is so small that the deformation is within the elastic range, the fuel rod may be reloaded in the reactor for further irradiation. (JIW)

  20. Theoretical requirements to tolerances to be imposed on fuel rod design parameters for RBEC-M lead-bismuth fast reactor

    International Nuclear Information System (INIS)

    Vasiliev, A.; Alekseev, P.; Mikityuk, K.; Fomichenko, P.; Shestopalov, A.

    2002-01-01

    Development of advanced reactors with innovative materials requires comprehensive analysis of fuel rod design parameters as well as tolerances to be imposed on these parameters. Currently, it is considered traditional to estimate uncertainties in core neutronics parameters on the basis of known tolerances imposed on fuel rod design parameters. However, requirements to some core neutronics parameters of advanced reactors can be first formulated and then taken into account, while developing the technologies for innovative fuel rod manufacturing, i.e. an 'inverse' problem can be solved. The aim of this problem is to find combinations of fuel rod design tolerances which provide that selected core neutronics parameters remain within specified deviations during base irradiation. (authors)

  1. Replacement rod

    International Nuclear Information System (INIS)

    Hatfield, S.C.

    1989-01-01

    This patent describes in an elongated replacement rod for use with fuel assemblies of the type having two end fittings connected by guide tubes with a plurality of rod and guide tube cell defining spacer grids containing rod support features and mixing vanes. The grids secured to the guide tubes in register between the end fittings at spaced intervals. The fuel rod comprising: an asymmetrically beveled tip; a shank portion having a straight centerline; and a permanently diverging portion between the tip and the shank portion

  2. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  3. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    International Nuclear Information System (INIS)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60% 235 U; the mini-rods were irradiated to an average burnup of ∼ 85% 235 U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%

  4. Cladding tube of fuel rod for a BWR type reactor

    International Nuclear Information System (INIS)

    Nakayama, Hitoshi; Fujie, Kunio; Kuwahara, Heikichi; Hirai, Tadamasa; Kakizaki, Kimio.

    1976-01-01

    Object: To form a cladding tube wall with tunnels in communication with the exterior through a number of small-diameter openings to rapidly disperse a large quantity of heat thereby providing high density of the fuel rod. Structure: Tunnels adjacent to each other are provided under the skin in contact with cooling liquid of a cladding tube, and a number of openings through which said tunnels and the periphery of the cladding tube are placed in communication are formed, said openings each having its section smaller than that of said tunnel. With this arrangement, the cooling water entered the tunnel through some of small diameter openings absorbs heat of the fuel rod to be vaporized, which is flown out into the cooling water through the other small diameter openings and formed into vapor bubbles which move up for release of heat. (Taniai, N.)

  5. 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.

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

    International Nuclear Information System (INIS)

    Lanthen, Jonas

    2006-09-01

    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

  7. Laser cutting equipment for dismantling irradiated PFR fuel sub-assemblies

    International Nuclear Information System (INIS)

    Higginson, P.R.; Campbell, D.A.

    1981-01-01

    Laser cutting was identified as a possible technique for dismantling irradiated Prototype Fast Reactor (P.F.R.) fuel sub-assemblies and initial trials showed that it could be used to make essentially swarf free cuts in P.F.R. wrapper material provided sufficient laser power was available to allow use of an inert cutting gas. A programme of development work has established a technique for inert gas cutting with the reliable, commercially available Ferranti MF 400 laser and equipment for laser cutting of sub-assemblies has been installed in the Irradiated Fuel Cave at P.F.R. Test cuts carried out with this equipment on un-irradiated wrapper sections have shown it to be easy to operate remotely, optically stable and reliable in operation. (author)

  8. 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

  9. Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

    Science.gov (United States)

    Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

    2018-01-01

    Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been

  10. Development of the vibration analysis technique of fuel rod and research on the methodology of fuel fretting wear analysis

    International Nuclear Information System (INIS)

    Kang, Heung Seok; Kim, Kyung Kyu; Yoon, Hyung Hoo; Song, Ki Nam

    1998-12-01

    The FEM program has been developed to predict the natural frequencies, the FEM program has been developed to predict the natural frequencies, and mode shapes of fuel rod subjected to axial force and continuously supported by a rotational and vent spring system, and to calculate the minimum reaction forces of the spacer grid spring when the maximum vibration amplitude of fuel rod is known. This program has been verified by commercial ANSYS program and the vibration test of dummy rods in air. The test equipment were set to get the fifth modes of test rods. Partial slip problem has been studied for the analysis of fuel fretting problem. Firstly, the assumption of semi-infiniteness of the contact bodies were validated by finite element (FE) analysis. From FE results, a classical bodies were validated by finite element (FE) analysis. From FE results, a classical theory of elasticity was utilized with regarding the problem as a plane problem. Secondly, the Mindlin-Cattaneo problem was re-evaluated, which gave the fundamental idea for developing the numerical tool for the shear traction on the contact. Shear force of sequentially-changing directions was considered and the corresponding shear traction was evaluated by extending the numerical tool for the Mindlin-Cattaneo problem

  11. Development of the vibration analysis technique of fuel rod and research on the methodology of fuel fretting wear analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Heung Seok; Kim, Kyung Kyu; Yoon, Hyung Hoo; Song, Ki Nam

    1998-12-01

    The FEM program has been developed to predict the natural frequencies, the FEM program has been developed to predict the natural frequencies, and mode shapes of fuel rod subjected to axial force and continuously supported by a rotational and vent spring system, and to calculate the minimum reaction forces of the spacer grid spring when the maximum vibration amplitude of fuel rod is known. This program has been verified by commercial ANSYS program and the vibration test of dummy rods in air. The test equipment were set to get the fifth modes of test rods. Partial slip problem has been studied for the analysis of fuel fretting problem. Firstly, the assumption of semi-infiniteness of the contact bodies were validated by finite element (FE) analysis. From FE results, a classical bodies were validated by finite element (FE) analysis. From FE results, aclassical theory of elasticity was utilized with regarding the problem as a plane problem. Secondly, the Mindlin-Cattaneo problem was re-evaluated, which gave the fundamental idea for developing the numerical tool for the shear traction on the contact. Shear force of sequentially-changing directions was considered and the corresponding shear traction was evaluated by extending the numerical tool for the Mindlin-Cattaneo problem.

  12. Band Width of Acoustic Resonance Frequency Relatively Natural Frequency of Fuel Rod Vibration

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich; Moukhine, V.S.; Novikov, K.S.; Galivets, E.Yu. [MPEI - TU, 14, Krasnokazarmennaya str., Moscow, 111250 (Russian Federation)

    2009-06-15

    In flow induced vibrations the fluid flow is the energy source that causes vibration. Acoustic resonance in piping may lead to severe problems due to over-stressing of components or significant losses of efficiency. Steady oscillatory flow in NPP primary loop can be induced by the pulsating flow introduced by reactor circulating pump or may be set up by self-excitation. Dynamic forces generated by the turbulent flow of coolant in reactor cores cause fuel rods (FR) and fuel assembly (FA) to vibrate. Flow-induced FR and FA vibrations can generally be broken into three groups: large amplitude 'resonance type' vibrations, which can cause immediate rod failure or severe damage to the rod and its support structure, middle amplitude 'within bandwidth of resonance frequency type' vibrations responsible for more gradual wear and fatigue at the contact surface between the fuel cladding and rod support and small amplitude vibrations, 'out of bandwidth of resonance frequency type' responsible for permissible wear and fatigue at the contact surface between the fuel cladding and rod support. Ultimately, these vibration types can result in a cladding breach, and therefore must be accounted for in the thermal hydraulic design of FR and FA and reactor internals. In paper the technique of definition of quality factor (Q) of acoustic contour of the coolant is presented. The value of Q defines a range of frequencies of acoustic fluctuations of the coolant within which the resonance of oscillations of the structure and the coolant is realized. Method of evaluation of so called band width (BW) of acoustic resonance frequency is worked out and presented in the paper. BW characterises the range of the frequency of coolant pressure oscillations within which the frequency of coolant pressure oscillations matches the fuel assembly's natural frequency of vibration (its resonance frequency). Paper show the way of detuning acoustic resonance from natural

  13. Thermalhydraulic behavior of electrically heated rod during a critical heat flux transient

    International Nuclear Information System (INIS)

    Lima, Rita de Cassia Fernandes de; Carajilescov, Pedro

    1997-01-01

    In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, a theoretical analysis of the drying and rewetting front propagation is performed during a critical heat flux experiment, starting with the application of slope of electrical power from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies done with several values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one. (author)

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

    International Nuclear Information System (INIS)

    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

  15. Fuel Rod Performance Evaluation of CE 16 x 16 LTA Operated at Steady State Using Transuranus and Pad Codes

    Energy Technology Data Exchange (ETDEWEB)

    Krasnorutskyy, V.; Slyeptsov, O. [Nuclear Fuel Cycle Science and Technology Establishment (NFCSTE), National Science Center, Kharkhov Institute of Physics and Technology (NSC KIPT), Kharkhov (Ukraine)

    2013-03-15

    The report performed under IAEA research contract No. 15370 describes the results of fuel performance evaluation of PWR fuel rods operated at steady state up to discharge burnup of {approx}60 GWD/MTU using the codes of TRANSURANUS designed by ITU and PAD designed by Westinghouse. The experimental results from US-PWR 16x16 LTA Extended Burnup Demonstration Program presented in the IFPE database of the OECD/NEA have been utilized for assessing the codes themselves during simulation of such properties as rod burnup, cladding corrosion, fuel densification and swelling, cladding irradiation growth and strain, FGR and RIP. The results obtained by PAD showed that the code properly simulates rod burnup, cladding irradiation growth and cladding oxidation with Standard Zr-4 material. The calculated burnup values along the fuel stack vary within {+-} 5% of the rod average burnup. The predicted values of the rod axial growth are (0.88-0.94) % and within the measured ones obtained in the burnup range of (50 - 60) GWD/MTU. With allowance made for probability of crud deposition and hot channel hydraulic diameter variation, the axial distribution of oxide layer is predicted well. For the nominal rod dimensions and operation conditions, the calculated peak oxide thickness is slightly overestimated based on the BE corrosion model parameters. The WEC fuel swelling and densification model together with the US NRC one, which is incorporated in the code, were used to assess the change in fuel pellet density ({Delta}{rho}) and fuel volume ({Delta}V{sub F}/V) vs. burnup as well as the rod void volume change, {Delta}V{sub V}/V, and the cladding outer diameter (OD) variation along the fuel stack. (author)

  16. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    International Nuclear Information System (INIS)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk; Kim, Hyochan; Yang, Yongsik

    2014-01-01

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  17. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Hyochan; Yang, Yongsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  18. Detection of leak-defective fuel rods using the circumferential Lamb waves excited by the resonance backscattering of ultrasonic pulses

    International Nuclear Information System (INIS)

    Choi, M.S.; Yang, M.S.; Kim, H.C.

    1992-01-01

    A new ultrasonic technique for detecting the infiltrated water in leaked fuel rods is developed. Propagation characteristics of the circumferential Lamb waves in the cladding tubes are estimated by the resonance scattering theory. The Lamb waves are excited by the resonance backscattering of ultrasonic pulses. In sound fuel rods, the existence of the Lamb waves is revealed by a series of periodic echoes. In leaked fuel rods, however, the Lamb waves are perturbed strongly by the scattered waves from the surface of fuel pellets, thus the periodic echoes are not observed. (author)

  19. CFD modeling of secondary flows in fuel rod bundles

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi

    2004-01-01

    An optimized non-linear eddy viscosity model is introduced, for calculations of detailed coolant velocity distribution in a tight lattice fuel bundle. The low Reynolds formulation has been optimized based on DNS data for channel flow. The non-linear stress-strain relationship has been modified in the coefficients to model the flow anisotropy, which causes the formation of turbulence driven secondary flows inside the bundle subchannels. Predictions of the model are first compared to experimental measurements of secondary flows in a triangularly arrayed rod bundle with p/d=1.3. Subsequently wall shear stress and velocity predictions are compared with different experimental data for a rod bundle with p/d=1.17. The model shows to be able to correctly reproduce the scale of the secondary motion, and to accurately reproduce both wall shear stress and velocity distributions inside the rod bundle subchannels. (author)

  20. Thermophysical instruments for non-destructive examination of tightness and internal gas pressure or irradiated power reactor fuel rods

    International Nuclear Information System (INIS)

    Pastoushin, V.V.; Novikov, A.Yu.; Bibilashvili, Yu.K.

    1998-01-01

    The developed thermophysical method and technical instruments for non-destructive leak-tightness and gas pressure inspection inside irradiated power reactor fuel rods and FAs under poolside and hot cell conditions are described. The method of gas pressure measuring based on the examination of parameters of thermal convection that aroused in gas volume of rod plenum by special technical instruments. The developed method and technique allows accurate value determination of not only one of the main critical rod parameters, namely total internal gas pressure, that forms rod mean life in the reactor core, but also the partial pressure of every main constituent of gaseous mixture inside irradiated fuel rod, that provides the feasibility of authentic and reliable leak-tightness detection. The described techniques were experimentally checked during the examination of all types power reactor fuel rods existing in Russia (WWER, BN, RBMK) and could form the basis for new technique development for non-destructive examination of PWR (and other) type rods and FAs having gas plenum filled with spring or another elements of design. (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)

    Likhanskii, V.; Evdokimov, I.; Khoruzhii, O.; Sorokin, A.; Novikov, V.

    2003-01-01

    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. Characteristics and properties of cladding tubes for VVER-1000 higher Uranium content fuel rods

    International Nuclear Information System (INIS)

    Peregud, M.; Markelov, A.; Novikov, V.; Gusev, A.; Konkov, V.; Pimenov, Y.; Agapitov, V.; Shtutsa, M.

    2009-01-01

    To improve the fuel cycle economics and to further increase the VVER fuel usability the work programme is under way to design novel improved fuel, fuel rods and fuel assemblies. Longer FA operation time that is needed to increase the fuel burnup and the related design developments of novel fuel assemblies resulted not only in changing types and sizes of Zirconium items and fuel assembly components but also altered the requirements placed on their technical characteristics. To use fuel rods having a larger charge of fuel, to improve their behaviour in LOCA, to reduce fuel rod damage ability during assembling the work was carried out to perfect the characteristics of both the cladding (reduced wall thickness and more rigid tolerances for geometry) and its material. To meet the more rigid requirements for the geometry dimensions of cladding tubes an improved process flow sheet has been designed and employed for their fabrication and also the finishing treatment of tube surfaces has been improved. The higher and stable properties of the cladding materials were managed through using the special purity in terms of Hafnium Zirconium (not higher than 100 ppm Hf) as a base of the E110 alloy and maintaining within the valid specifications for the alloy the optimized contents of Oxygen and Iron at the levels of (600 - 990) ppm and (250 - 700) ppm, respectively. The work was under way in 2004 - 2008 years; during this period the technology and materials science solutions were mastered that were phased-in introduced into the production of the cladding tubes for the fuels loaded into the of the Kalinin NPP Unit 1

  3. Plugger guide for aligning an end plug and a fuel rod tube end

    International Nuclear Information System (INIS)

    Klapper, K.K.; Boatwright, D.A.

    1987-01-01

    A pin driving tool is described for inserting or removing pins from teeth on a digging means, comprising: fuel rod tube toward an end plug for application of the end plug into the tube end, the apparatus comprising: (a) a guide housing having an elongated central longitudinal bore with one end for receiving the end plug and an opposite end for receiving the fuel rod tube end; (b) sets of rolling elements disposed in the housing at axially spaced positions along and about the bore thereof. The rolling elements in each set are positioned in fixed relation with respect to one another to receive the fuel rod tube end therebetween and align the tube end with the end plug as the tube end is moved through the bore and into engagement with the end plug; and (c) retaining means disposed adjacent to the open end of the housing bore for engaging the end plug so as to maintain it in a stationary seated position at the one end of the housing bore

  4. Core conversion from rod to plate type fuel elements in research reactors

    International Nuclear Information System (INIS)

    Khattab, M.S.; Mina, A.R.

    1997-01-01

    Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

  5. The measurements of critical mass with uranium fuel elements and thorium rods

    International Nuclear Information System (INIS)

    Yao Zhiquan; Chen Zhicheng; Yao Zewu; Ji Huaxiang; Bao Borong; Zhang Jiahua

    1991-01-01

    The critical experiments with uranium elements and Thorium rods have been performed in zero power reactor at Shanghai Institute of Nuclear Research. The critical masses have been measured in various U/Th ratios. The fuels are 3% 235 U-enriched uranium. The Thorium rods are made from power of ThF 4 . Ratios of calculated values to experimental values are nearly constant at 0.995

  6. Vibration analysis of a dummy fuel rod continuously supported by spacer grids

    International Nuclear Information System (INIS)

    Choi, Myoung-Hwan; Kang, Heung-Seok; Yoon, Kyung-Ho; Song, Kee-Nam; Jung, Youn-Ho

    2003-01-01

    A modal testing and a finite element (FE) analysis using ABAQUS on a dummy fuel rod continuously supported by Optimized H type (OHT) and New Doublet (ND) spacer grids are performed to obtain the vibration characteristics such as natural frequencies and mode shapes and to verify the FE model used. The results from the test and the FE analysis are compared according to modal assurance criteria values. The natural frequency differences between the two methods as well as the mode comparison results for the rod with OHT SG are better than those with ND SG. That is, in the case of the ND grid model using beam-spring elements, there was a large discrepancy between the two methods. Thus, we tried to modify the FE model for ND SG considering the contact phenomena between the fuel rod and the SG. The results of the new model showed good agreement with the experiment compared with those of a beam-spring model

  7. 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)

  8. The physical and chemical degradation of PWR fuel rods in severe accident conditions

    International Nuclear Information System (INIS)

    Parsons, P.D.; Mowat, J.A.S.; Dewhurst, D.W.F.; Hughes, T.E.

    1983-01-01

    An experimental study of the interaction between Zircaloy-4 cladding and UO 2 in PWR fuel rods heated to high temperatures with a negligible differential pressure across the cladding wall is described. The fuel rods were of dimensions appropriate to the 17x17 PWR fuel sub-assembly and were heated in a non-oxidising environment (vacuum) up to approx. 1850 deg. C either isothermally or through heating ramps. Observations were made concerning the extent and nature of the reaction zone between Zircaloy-4 and UO 2 over the temperature range 1500-1850 deg. C for times ranging from 1 min to 125 min. The location, morphology and the chemical composition of the phases formed are described along with the kinetics of their formation. (author)

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

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Abe, Tomoyuki

    2006-03-01

    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 UO 2 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

  10. Status of rod consolidation, 1988

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1989-01-01

    It is estimated that the spent fuel storage pools at some domestic light-water reactors will run out of space before 2003, the year that the US Department of Energy currently predicts it will have a repository available. Of the methods being studied to alleviate the problem, rod consolidation is one of the leading candidates for achieving more efficient use of existing space in spent fuel storage pools. Rod consolidation involves mechanically removing all the fuel rods from the fuel assembly hardware (i.e., the structural components) and placing the fuel rods in a close-packed array in a canister without space grids. A typical goal of rod consolidation systems is to insert the fuel rods from two fuel assemblies into a canister that has the same exterior dimensions as one standard fuel assembly (i.e., to achieve a consolidation or compaction ratio of 2:1) and to compact the nonfuel-bearing structural components from those two fuel assemblies by a factor of 10 to 20. This report provides an overview of the current status of rod consolidation in the United States and a small amount of information on related activities in other countries. 85 refs., 36 figs., 5 tabs

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

    International Nuclear Information System (INIS)

    2010-10-01

    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

  12. Review of Current Criteria of Spent Fuel Rod Integrity during Dry Storage

    International Nuclear Information System (INIS)

    Yang, Yong Sik; Kim, Sun Ki; Bang, Je Geon; Song, Kun Woo

    2006-01-01

    A PWR spent fuel has been stored in a wet storage pool in Korea. However, the amount of spent fuel is expected to exceed the capacity of a wet storage pool within 10∼15 years. From the early 1970's, a research on the PWR spent fuel dry storage started because the dry storage system has been economical compared with the wet storage system. The dry storage technology for Zircaloy-clad fuel was assessed and licensed in many countries such as USA, Canada, FRG and Switzerland. In the dry storage system, a clad temperature may be higher than in the wet storage system and can reach up to 400 .deg.. A higher clad temperature can cause cladding failures during the period of dry storage, and thus a dry storage related research has essentially dealt with the prevention of clad degradation. It is temperature and rod internal pressure that cause cladding failures through the mechanisms such as clad creep rupture, hydride re-orientation, and stress-corrosion cracking etc.. In this paper, the current licensing criteria are summarized for the PWR spent fuel dry storage system, especially on spent fuel rod integrity. And it is investigated that an application propriety of existing criteria to Korea spent fuel dry storage system

  13. 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)

    Horwood, W.A.

    1981-07-01

    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 90 0 included angle. Data is provided on the effectiveness of the manufacturing operations chosen for grid machining and assembly

  14. Critical heat flux detection in rods simulating fuel elements by using dilation method

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1993-01-01

    In out-reactor heat transfer experiments, fuel elements are often simulated by electrically heated rods. In order to prevent the heating rod from being damaged by burnout, when the critical heat flux occurs a safety system is provided which checks the axial thermal expansion of the rod. In case of sudden temperature increase, the corresponding elongation causes a fast interruption of the electrical power supply. The experiments presented here show that this method is more effective than one that uses thermocouples. (author)

  15. Vernotte-Cattaneo approximation for heat conduction in fuel rod

    International Nuclear Information System (INIS)

    Espinosa P, G.; Espinosa M, E. G.

    2009-10-01

    In this paper we explore the applicability of a fuel rod mathematical model based on the Vernotte-Cattaneo transient heat conduction as constitutive law (Non-Fourier approach) for light water reactors transient analysis. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The motivation for this research was to eliminate the paradox of an infinite. The motivation for this research was to eliminate the paradox of an infinite thermal wave speed. The time-dependent heat sources were considered in the fuel rod heat transfer model. The close of the main steam isolated valves transient in a boiling water reactor was analyzed for different relaxation times. The results show that for long-times the heat fluxes on the clad surface under Vernotte-Cattaneo approach can be important, while for short-times and from the engineering point of view the changes are very small. (Author)

  16. Operational experience gained with the failed fuel rod detection system in nuclear power plants

    International Nuclear Information System (INIS)

    Boehm, H.H.; Forch, H.

    1985-01-01

    Brown Boveri Reaktor GmbH together with Krautkramer Company developed such a FAILED FUEL ROD DETECTION SYSTEM (FFRDS) which allows to located defective fuel rods without dismantling the fuel assembly or pulling of individual rods. Since 1979 the FFRDS is employed successfully in various nuclear power plants in Europe, USA, Japan, and Korea. The short inspection time and the high reliability of the method make the FFRDS a true competitor to the sipping method. In this paper the authors discuss the method and the design of the system, the equipment set-up, its features and the experience gained so far. The system has been performed and automated to such an extent that within a short installation period series of fuel assemblies can be tested with relatively short intervals of time (5 minutes for BWR and 7 minutes for PWR fuel assemblies per side). The ability of the system for deployment under various conditions and the experience gained during the past six years have made this system universally applicable and highly sensitive to the requirements of NDT during outages and for transport of FAs to intermediate storage facilities. Comparison of FFRDS to conventional sipping has indicated in several instances that the FFRDS is superior to the latter technique

  17. Thermal and stress analysis of a fuel rod research project 277

    International Nuclear Information System (INIS)

    1975-04-01

    The purpose of this investigation was to perform an analytical evaluation of a postulated loss of coolant incident in a large pressurized water reactor. A coupled thermal and stress finite element analysis of a fuel rod subjected to a hypothetical blow-down transient was carried out. The effect of two gap conditions and two initial stress states on the response of the fuel rod was studied. Both one-dimensional and three-dimensional models were investigated. To study the heat transfer in the gap region one assumes a conductive mode of heat transfer in the gap characterized by an equivalent thermal conductivity, which is dependent on the current gap width. Accordingly, coupled analysis procedure and computational scheme were established. A mesh generation computer program was developed for the three-dimensional model

  18. Post-irradiation examination of A1-61 wt % U3Si fuel rods from the NRU reactor

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    1997-09-01

    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 D 2 0 coolant inlet temperature 37 degrees 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 % 235 U 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 μm thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 μm 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 A1-61 wt % U 3 Si fuel irradiated in the NRU reactor. (author)

  19. Parametric study of fuel rod behaviour during the RIA using the modified FALCON code

    International Nuclear Information System (INIS)

    Khvostov, G.; Zimmermann, M.A.; Ledergerber, G.

    2010-01-01

    Presented in the paper are the results of a parametric study with the use of optimised modules of the FALCON code (FALCON-PSI) that addresses the effects of the selected characteristics of fast thermal transients (e.g., impulse width), fuel rod design (e.g., active fuel attack length) and boundary conditions (e.g., the coolant conditions) on fuel behaviour during a RIA. Specifically, the analysis of the governing processes for the fuel rod behaviour during the RIA events simulated in the experimental facility of the Nuclear Safety Research Reactor (NSRR, Japan) are in the focus of the present study. The results obtained can be useful for a better transfer of the NSRR test results in relation to the corresponding behaviour in LWRs and furthermore might also support the planning of future additional experiments. (authors)

  20. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified

    International Nuclear Information System (INIS)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes

    2014-01-01

    This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results

  1. An evaluation of the influence of fuel design parameters and burnup on pellet/cladding interaction for boiling water reactor fuel rod through in-core diameter measurement

    International Nuclear Information System (INIS)

    Yanagisawa, K.

    1986-01-01

    The influence of design parameters and burning on pellet/cladding interaction (PCI) of current boiling water reactor fuel rods was studied through in-core diameter measurement. Thinner cladding and a smaller diametral gap enhanced the PCI during startup. At constant power, fuel with SiO 2 added greatly reduced PCI due to relaxation. The fuel with a small grain size greatly reduced PCI due to densification. Preirradiation of rods up to 23 MWd/kgU caused a large PCI not only in a small gap but also in a large gap rod. Relaxation and permanent deformation was small. In the power increase experiment, one rod experienced PCI failure. The spurt times of coolant radioactivity coincided well with the sudden drop of cladding axial strain and marked crack opening at the rod surface. The estimated hoop stress predicted by FEMAXI-III was 350 MPa at the failure

  2. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...

  3. Fuel Rod Consolidation Project: Phase 2, Final report: Volume 2, Appendices

    International Nuclear Information System (INIS)

    1987-01-01

    This document, Volume 2, provides the appendices to Volume 1 of the Fuel Rod Consolidation Project. It provides information on the following: References; Trade-off Studies; Instrument List; RAM Data; Fabrication Specifications; Software Specifications; and Design Requirements

  4. Computerized operating procedures for shearing and dissolution of segments from LWBR [Light Water Breeder Reactor] fuel rods

    International Nuclear Information System (INIS)

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work

  5. Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

    International Nuclear Information System (INIS)

    Han, Kyu Hyun; Chang, Soon Heung

    2005-01-01

    More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

  6. Nuclear fuel rod with burnable plate and pellet-clad interaction fix

    International Nuclear Information System (INIS)

    Boyle, R.F.

    1987-01-01

    This patent describes a nuclear fuel rod comprising a metallic tubular cladding containing nuclear fuel pellets, the pellets containing enriched uranium-235. The improvement described here comprises: ceramic wafers, each wafter comprising a sintered mixture of gadolinium oxide and uranium dioxide, the uranium oxide having no more uranium-235 than is present in natural uranium dioxide. Each of the wafers is axially disposed between a major portion of adjacent the nuclear fuel pellets, whereby the wafers freeze out volatile fission products produced by the nuclear fuel and prevent interaction of the fission products with the metallic tubing cladding

  7. Development of oxygen sensing technology in an irradiated fuel rod. Characteristic test of oxygen sensor

    International Nuclear Information System (INIS)

    Saito, Junichi; Hoshiya, Taiji; Sakurai, Fumio; Sakai, Haruyuki

    1996-03-01

    At the Department of JMTR (Japan Materials Test Reactor), the re-instrumentation technologies to a high burnup fuel rod irradiated in an LWR have been developed to study irradiation behavior of the fuel during power transient. It has been progressed developing a chemical sensor as one of the re-instrumentation technologies. This report summarizes the results of characteristic tests of an oxygen sensor made of Yttria Stabilized Zirconia (YSZ) as a solid electrolyte. Several kinds of experiments were carried out to evaluate the electromotive force (emf) performance, stability and lifetime of the oxygen sensor with Ni/NiO, Cr/Cr 2 O 3 and Fe/FeO, respectively as a reference electrode. From the experimental data, it is suggested that the reference electrode of Ni/NiO reveals the most appropriate characteristic of the sensor to measure the partial oxygen pressure in a fuel rod. It is the final goal of this development to clarify the change of oxygen chemical potential in a fuel rod during power transient. (author)

  8. Development of a new bench for puncturing of irradiated fuel rods in STAR hot laboratory

    Science.gov (United States)

    Petitprez, B.; Silvestre, P.; Valenza, P.; Boulore, A.; David, T.

    2018-01-01

    A new device for puncturing of irradiated fuel rods in commercial power plants has been designed by Fuel Research Department of CEA Cadarache in order to provide experimental data of high precision on fuel pins with various designs. It will replace the current set-up that has been used since 1998 in hot cell 2 of STAR facility with more than 200 rod puncturing experiments. Based on this consistent experimental feedback, the heavy-duty technique of rod perforation by clad punching has been preserved for the new bench. The method of double expansion of rod gases is also retained since it allows upgrading the confidence interval of volumetric results obtained from rod puncturing. Furthermore, many evolutions have been introduced in the new design in order to improve its reliability, to make the maintenance easier by remote handling and to reduce experimental uncertainties. Tightness components have been studied with Sealing Laboratory Maestral at Pierrelatte so as to make them able to work under mixed pressure conditions (from vacuum at 10-5 mbar up to pressure at 50 bars) and to lengthen their lifetime under permanent gamma irradiation in hot cell. Bench ergonomics has been optimized to make its operating by remote handling easier and to secure the critical phases of a puncturing experiment. A high pressure gas line equipped with high precision pressure sensors out of cell can be connected to the bench in cell for calibration purposes. Uncertainty analyses using Monte Carlo calculations have been performed in order to optimize capacity of the different volumes of the apparatus according to volumetric characteristics of the rod to be punctured. At last this device is composed of independent modules which allow puncturing fuel pins out of different geometries (PWR, BWR, VVER). After leak tests of the device and remote handling simulation in a mock-up cell, several punctures of calibrated specimens have been performed in 2016. The bench will be implemented soon in hot

  9. Construction of supporting grids for fuel rods (or tubes in a heat exchanger)

    International Nuclear Information System (INIS)

    1975-01-01

    The construction of supporting grids for fuel rods (or tubes in heat exchangers) is described. It is a modification of a former French patent. The modification consists in the use of different material for the springs keeping the rod in place and describes another way of fixing these blade-shaped springs. Advantages of the specific spring characteristics were taken into consideration

  10. Water rod and fuel assembly

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Tada, Nobuo; Nakajima, Junjiro; Aizawa, Yasuhiro.

    1995-01-01

    A water rod disposed in a fuel assembly comprises a larger diameter tube constituting an upwarding flow channel for coolants flown from the lower portion of a reactor core, and a smaller diameter tube connected fixedly to the larger diameter tube at the periphery of the upper end thereof and constituting a downwarding flow channel for coolants upwardly flown in the larger diameter tube. The larger diameter tube is formed by subjecting a base tube made of a zirconium alloy to PILGER mil fabrication and annealing in α region repeatingly for several times, then subjecting it to α + β treatment for once. The smaller diameter tube is formed by subjecting a base tube made of a zirconium alloy to PILGER mil fabrication and annealing in α region repeatingly for several times, then subjecting it to β treatment for once. With such procedures, the amount of irradiation growth of the tube in the axial direction is made greater in the larger diameter tube than that in the smaller diameter tube. Accordingly, since the smaller diameter tube is never bent by pressing, mechanical integrity of the fuel assembly is never lost. (I.N.)

  11. Evaluation of fuel rod damage in LWR under accident conditions using SSYST

    International Nuclear Information System (INIS)

    Meyder, R.

    1982-01-01

    After a short outline of the recent SSYST-development, the creep rupture model NORA 2 is presented. The effect of temperature and oxygen on Zircaloy 4 creep behaviour is shown. Examples on the effect of azimuthal varying gap width and wall thickness are given. Remarks on the extension of a single rod analysis on a bundle and the stepwise application of SSYST for investigation of fuel rod failure conclude the paper. (orig.) [de

  12. Apparatus for adjusting the elevation of fuel rods in a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Hale, D.L.; Culbreth, T.F.

    1988-01-01

    A tool adapted for adjusting the level of a nuclear fuel rod in a fuel assembly is described comprising: an expander comprising two elongate generally parallel and laterally spaced apart arms extending in a longitudinal direction, an actuator operatively mounted to the expander so as to be disposed between the two arms and being movable with respect thereto, and cooperating surface means mounted to the arms and the actuator for laterally separating the free ends of the arms to a predetermined maximum distance upon movement of the actuator with respect to the arms

  13. Out-of-pile experiments on the high-temperature behavior of Zircaloy-4 clad fuel rods

    International Nuclear Information System (INIS)

    Hagen, S.

    1984-01-01

    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/uranium dioxide (UO 2 ) fuel rod simulators surrounded with a Zircaloy shroud--simulating the Zircaloy of neighboring 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 2200 0 C. 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 behavior of the shroud, which was itself affected by the insulation and steam cooling. Damage to the fuel rods increased with increasing heatup rate. Fro slow heatup rates nearly no interaction between the oxidized cladding and UO 2 was observed, while for fast heatup rates the entire annular pellet was dissolved by molten Zircaloy

  14. Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

    Science.gov (United States)

    Muratov, V. G.; Lopatkin, A. V.

    An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

  15. Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

    International Nuclear Information System (INIS)

    Caraghiaur, Diana; Anglart, Henryk; Frid, Wiktor

    2009-01-01

    This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.

  16. Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Caraghiaur, Diana [Royal Institute of Technology, Division of Nuclear Reactor Technology, Department of Physics, School of Engineering Sciences, AlbaNova University Center, SE-106 91 Stockholm (Sweden)], E-mail: dianac@kth.se; Anglart, Henryk [Royal Institute of Technology, Division of Nuclear Reactor Technology, Department of Physics, School of Engineering Sciences, AlbaNova University Center, SE-106 91 Stockholm (Sweden); Frid, Wiktor [Swedish Radiation Safety Authority, Reactor Technology and Structural Integrity, SE-171 16 Stockholm (Sweden)

    2009-10-15

    This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.

  17. Study of corium radial spreading between fuel rods in a PWR core

    International Nuclear Information System (INIS)

    Roche, S.; Gatt, J.M.

    1996-01-01

    In the framework of severe accident studies for PWR like Three Mile Island Unit 2 (TMI-2), the reactor core essentially constituted of fuel rods begins to heat and then to melt. During the early degradation phase, a melt (essentially UO2 and ZrO2) that constitutes the corium flows first along the rods, and after a blockage formation, may radially propagate towards the core periphery. A simplified model has been elaborated to study the corium freezing phenomena during its crossflow between the fuel rods. The corium spreads on an horizontal support made, of either a corium crust, or a grid assembly. The model solves numerically the interface energy balance equation at the solid-liquid corium interface and the monodimensional heat balance equation in transient process with convective terms and heat source (residual power). ''Zukauskas'' correlations are used to calculate heat transfer coefficients. The model can be integrated in severe accident codes like ICARE II (IPSN) describing the in-vessel degradation scenarios. (author). 5 refs, 10 figs

  18. Modeling and preliminary analysis on the temperature profile of the (TRU-Zr)-Zr dispersion fuel rod for HYPER

    International Nuclear Information System (INIS)

    Lee, B. W.; Hwang, W.; Lee, B. S.; Park, W. S.

    2000-01-01

    Either TRU-Zr metal alloy or (TRU-Zr)-Zr dispersion fuel is considered as a blanket fuel for HYPER(Hybrid Power Extraction Reactor). In order to develop the code for dispersion fuel rod performance analysis under steady state condition, the fuel temperature distribution model which is the one of the most important factors in a fuel performance code has been developed in this paper,. This developed model computes the one dimensional radial temperature distribution of a cylindrical fuel rod. The temperature profile results by this model are compared with the temperature distributions of U 3 Si-A1 dispersion fuel and TRU-Zr metal alloy fuel. This model will be installed in performance analysis code for dispersion fuel

  19. Heat transfer coefficient testing in nuclear fuel rod bundles with mixing vane grids

    International Nuclear Information System (INIS)

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2005-01-01

    An air heat transfer test facility was developed to test the heat transfer downstream of support grids in simulated PWR nuclear fuel rod bundles. The goal of this testing is to study the single-phase heat transfer coefficients downstream of grids with mixing vanes in a square-pitch rod bundle. The technique developed utilizes fully-heated grid spans and a specially designed thermocouple holder that can be moved axially down the rod bundle and aximuthally within a test rod. From this testing, the axial and aximuthally varying heat transfer coefficient can be determined. Different grid designs are tested and compared to determine the heat transfer enhancement associated with key grid features such as mixing vanes. (author)

  20. ROBOT3: a computer program to calculate the in-pile three-dimensional bowing of cylindrical fuel rods (AWBA Development Program)

    International Nuclear Information System (INIS)

    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